de maximis, inc., inc.while it is in fact possible that some disposed uranium could resist the hot...

66
de maximis, inc., inc. 200 Day Hill Road Suite 200 Windsor, CT 06095 (860) 298-0541 (860) 298-0561 FAX April 3, 2006 Ms. Melissa Taylor US EPA, Region 1 OSRR, MA Superfund Section One Congress Street, Suite 1100 Boston, MA 02114-2023 Subject: Response to Comments in CREW Technical Note No. 2006-01 Dear Ms. Taylor: We have reviewed the comments provided in the Citizens Research & Environmental Watch (CREW) Technical Note No. 2006-01 submitted to EPA January 5, 2006 (CREW, 2006) regarding their disagreement with some of the conclusions presented in the Technical Memorandum “Radiological Review of Fall 2004 RI Analytical Data, Nuclear Metals Superfund Site” (the Technical Memorandum), September 26, 2005 (MACTEC, 2005). This letter provides responses to the information presented in the Technical Note, and re-iterates the intention and rationale for discontinuing the use of alpha spectroscopy for the analysis of uranium and thorium isotopes at the Nuclear Metals, Inc. Superfund Site (the Site). Introduction The Technical Memorandum presents an evaluation of radionuclide data for various media collected during the fall 2004 sampling program conducted at the Site. This evaluation concludes the uranium isotopic profile at the Site consists of depleted and natural uranium, and the thorium isotopic profile consists of thorium-232. In addition to verifying the uranium and thorium isotopic profiles, the evaluation also demonstrates the lack of enriched uranium or hard to detect radionuclides (HTDRs) at the Site as a result of past activities. In light of the findings presented in this evaluation, the Technical Memorandum concludes uranium and thorium are the only radionuclides of concern in the evaluation of the nature and extent of contamination and alpha spectroscopy is not needed for future analysis of uranium and thorium isotopes at the Site. As discussed below, this letter will clarify the intention of the Technical Memorandum and rationale for discontinuing the use of alpha spectroscopy analysis for future samples collected at the Site, as well as provides responses to comments presented in the CREW Technical Note. In particular, the following points will be discussed in greater detail in sections that follow: Allentown, PA – Clinton, NJ – Greensboro, GA – Knoxville, TN – Farmington Hills, MI – Riverside, CA Cortland, NY – St. Charles, IL – Sarasota, FL – Jacksonville, FL – Houston, TX – Windsor, CT – Waltham, MA

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Page 1: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

de maximis inc inc 200 Day Hill Road

Suite 200 Windsor CT 06095

(860) 298-0541 (860) 298-0561 FAX

April 3 2006

Ms Melissa Taylor US EPA Region 1 OSRR MA Superfund Section One Congress Street Suite 1100 Boston MA 02114-2023

Subject Response to Comments in CREW Technical Note No 2006-01

Dear Ms Taylor

We have reviewed the comments provided in the Citizens Research amp Environmental Watch (CREW) Technical Note No 2006-01 submitted to EPA January 5 2006 (CREW 2006) regarding their disagreement with some of the conclusions presented in the Technical Memorandum ldquoRadiological Review of Fall 2004 RI Analytical Data Nuclear Metals Superfund Siterdquo (the Technical Memorandum) September 26 2005 (MACTEC 2005) This letter provides responses to the information presented in the Technical Note and re-iterates the intention and rationale for discontinuing the use of alpha spectroscopy for the analysis of uranium and thorium isotopes at the Nuclear Metals Inc Superfund Site (the Site)

Introduction

The Technical Memorandum presents an evaluation of radionuclide data for various media collected during the fall 2004 sampling program conducted at the Site This evaluation concludes the uranium isotopic profile at the Site consists of depleted and natural uranium and the thorium isotopic profile consists of thorium-232 In addition to verifying the uranium and thorium isotopic profiles the evaluation also demonstrates the lack of enriched uranium or hard to detect radionuclides (HTDRs) at the Site as a result of past activities In light of the findings presented in this evaluation the Technical Memorandum concludes uranium and thorium are the only radionuclides of concern in the evaluation of the nature and extent of contamination and alpha spectroscopy is not needed for future analysis of uranium and thorium isotopes at the Site

As discussed below this letter will clarify the intention of the Technical Memorandum and rationale for discontinuing the use of alpha spectroscopy analysis for future samples collected at the Site as well as provides responses to comments presented in the CREW Technical Note In particular the following points will be discussed in greater detail in sections that follow

Allentown PA ndash Clinton NJ ndash Greensboro GA ndash Knoxville TN ndash Farmington Hills MI ndash Riverside CA Cortland NY ndash St Charles IL ndash Sarasota FL ndash Jacksonville FL ndash Houston TX ndash Windsor CT ndash Waltham MA

de maximis inc

bull The fall 2004 radionuclide data supports the conclusion that future alpha spectroscopy analysis is not needed as the isotopic signature for both uranium and thorium has been defined and the lack of enriched uranium was demonstrated The purpose for the 5 (1 alpha spectroscopy20 ICPMS) alpha spectroscopy split analysis was to verify the original assumptions made in RIFS work plan (PSOP 2005) with respect to uranium and thorium isotopic profiles verify the uranium profile resembles natural or depleted isotopic mix and determine the thorium profile in site media so that the contribution of thorium can be distinguished from the background profile and that associated with uranium decay

bull Uranium and thorium results for ICPMS tend to be lower as compared to alpha spectroscopy analysis however this is not unexpected as ICPMS results are representative of environmental conditions at the site whereas alpha spectroscopy analysis results are representative of environmental and natural conditions combined

bull ICPMS uranium results for a small sub-set of samples are an analytical anomaly however this fact does not cast doubt on the entire ICPMS data set

bull The variability in results does not cast doubt on the reliability of the laboratory results

bull The nature and extent of uranium at the Site as well as clean-up levels will be based on the mass of uranium present not enrichment

bull Samples for HTDR analysis were collected as specified in the RIFS Work Plan (FSP 2004) and several of these samples were collected in areas with elevated uranium concentrations Further testing for these radionuclides is not needed

bull Uranium-232 and Uranium-236 measurements are not needed as HTDR analysis demonstrated the lack of reactor-produced radionuclides

bull Radium measurements are not needed as there was no historical use at the site and the historical result appears to be an anomaly

Response to Section 2 - Soil and sediment analysis by ICPMS versus alpha spectroscopy

The lsquosuper-depletedrsquo uranium points shown in Figure 1 of the CREW Technical Note (CREW 2006) are an anomaly in the ICPMS analysis and lsquosuper-depletedrsquo uranium does not exist However given the large amount of ICPMS uranium data points collected from the site (651 uranium data points for soil and sediment from fall 2004 data set) it is reasonable and not unexpected to find a few outliers in every sampling event Upon further investigation it was determined the analytical results for these data points are from two (2) laboratory sample delivery groups (SDGs) Although results for

2

de maximis inc

these samples have undergone data validation procedures no analytical or transcription errors were observed Furthermore even though anomalies are identified results from all other samples (including other samples within the same SDGs) are valid and usable These limited anomalies should not cast doubt on all other uranium data generated using ICPMS

A detailed review was performed on the uranium results for the lsquosuper-depletedrsquo samples The review identified the following however no specific reason for the anomaly was identified

bull Sample results are limited to only 2 SDGs for ICPMS

bull U-235 and U-238 isotopes were reported from the same analytical run

bull Calculated enrichment values (ratio U-235U-238) from ICPMS instrument printshyouts (expressed as ugL) matched calculated enrichment values from mass-converted results (expressed as mgkg)

bull No transcription errors were found sample calculations and dilutions were verified as compared to Report of Analysis Forms

bull Calibration curve slopes for each isotope were very similar (U-235 = 000007084 and U-238 = 000007368)

bull Laboratory preparation and batching worksheets showed that although samples were prepped batched and analyzed at similar times no distinct pattern was recognized that could lead to an explanation for errors (ie other samples were prepped batched and analyzed with the lsquosuper-depletedrsquo samples and results for these samples showed either depleted or natural isotopic profiles)

bull ICPMS uranium concentrations for these two SDGs were plotted (See Figure 1) in ascending order Note the discrete inflection point for sample concentrations starting at approximately 100 mgkg These points (plus 2 additional points greater than 1600 mgkg) correspond to the samples termed lsquosuper-depletedrsquo

To better assess the uranium concentrations at the ldquosuper-depletedrsquo sampling locations the lsquosuper-depletedrsquo samples will be re-collected as part of the Phase IC soil investigation

As demonstrated in the Technical Memorandum and as supported by an independent evaluation of uranium split-sample results prepared by Metcalf and Eddy (MampE) on behalf of EPA (refer to Attachment 1) the alpha spectroscopy analysis results generally provide higher results as compared to ICPMS analysis As discussed below and reshyiterated in the MampE evaluation the sample preparation and digestion techniques for the two methods are different as is sample quantitation

3

de maximis inc

The ICPMS digestion procedure is designed to evaluate metals that could become ldquoenvironmentallyrdquo available whereas the alpha spectroscopy digestion is an extremely aggressive technique where complete mineral dissolution is performed As a result the generally higher uranium concentrations reported for the alpha spectroscopy analyses are not unexpected and may be explained by the liberation of geologically-bound uranium during digestion

The following information describes the methodologies and differences between the two analytical methods

bull Preparation homogenization and digestion ndash laboratory SOPs for the RIFS (QAPP 2004) are presented in Attachment 2 for these techniques The alpha spectroscopy preparation technique (SOP GL-RAD-015) was adapted from the DOE HASL Manual and uses a larger aliquot of sample and dries grinds ashes the sample prior to acid digestion The alpha spectroscopy analysis uses complete dissolution (HNO3 and HF) The ICPMS technique (SOP GL-MA-Eshy009) follows the EPA digestion method approved for the RIFS (QAPP 2004) and includes homogenization of the wet sample (smaller aliquot) and digestion using hot acid (HNO3 and H2O2)

bull Sample Quantitation - alpha spectroscopy method quantitation procedure includes adjustment of instrument results based on tracer R (percent yield is factored into reported result) ICPMS method quantitation procedure does not include applying correction factors for any bias observed during analysis (ie an absolute value is reported)

bull The ICPMS hot acid digestion technique used for the RIFS is a standard EPA digestion technique used at hundreds of CERCLA sites and is a widely accepted and aggressive method that is appropriate for a wide range of metals including uranium and thorium Furthermore as documented in the laboratory ICPMS SOP (GL-MA-E-014) uranium isotopes and thorium are suitable metals for analysis

bull As shown in Attachment 1 there is a very strong comparison between uranium split-sample results reported by de maximis inc and MampE which demonstrates the representativeness and usability of ICPMS data for the Site

Another factor to consider as part of the comparison of analytical methods is the resolution of uranium by each method The primary isotope of natural and depleted uranium is U-238 which has a low specific activity (ie it takes a large amount of mass to increase the amount of radioactivity) For example each increase in radioactivity by 1 pCig of U-238 results in an increase in mass of 3 mgkg Due to the fact that ICPMS has better resolution and sensitivity at lower mass concentrations it was chosen (and approved by EPA) as the quantitative analytical method for uranium for the RIFS

4

de maximis inc

EPA metals digestion procedures for all analyses (whether ICP ICPMS cold vapor) do not ldquounder-leachrdquo samples rather digestion procedures are designed to evaluate ldquoenvironmentally availablerdquo concentrations As such uranium and thorium would be analyzed as for other potential metal contaminants at the site (Cu Be Pb Hg etc)

While it is in fact possible that some disposed uranium could resist the hot nitric digestion it would also certainly resist leaching by rainwater infiltration or groundwater flushing and be less biologically available if ingested

To assess laboratory performance of the QAPP-approved methods for the analysis of uranium and thorium for the project performance evaluation (PE) samples have been submitted to the laboratory As part of the Phase IA and Phase 1B investigation programs blind PE samples (quality control standard that is of a composition and concentration not known to the laboratory) were submitted to the laboratory for analysis Results of the analyses and comparison to vendor spike concentrations are provided below

Investigation Analysis Parameter Laboratory Result Certified Value (Vendor Spike)

QC PALs

Phase 1A ICPMS Uranium (total)

32 26 24 mgkg1 22 mgkg NA

Phase 1A ICPMS Thorium 229 159 162 110 mgkg1

131 mgkg NA

Phase 1A Alpha Spec

Uraniumshy238

313 pCig 339 pCig NA

Phase 1A Alpha Spec

Uraniumshy234

34 pCig 339 pCig NA

Phase 1A Alpha Spec

Uraniumshy235

0252 U pCig2 016 pCig NA

Phase 1A Alpha Spec

Thorium-230 261 pCig 134 pCig NA

Phase 1B Alpha Spec

Uranium (total)

1184 pCiL 117 pCiL 852 ndash 138 pCiL

Phase 1B ICPMS Uranium 95 mgkg 126 mgkg 834 ndash 144 mgkg

Phase 1B ICPMS Uranium 606 ugL 635 ugL 525 ndash 745 ugL

1 = results provided for multiple runs performed by the laboratory 2 = not detected laboratory minimum detectable activity is greater than amount present in PE sample NA = not available

As seen in this table the laboratory provided acceptable results (ie results generally showed good agreement to vendor certified values and results were reported within QC Performance Acceptance Limits of the vendor result (when applicable) as determined by peer laboratory measurements) indicating the QAPP methods used for uranium and thorium analysis are capable of providing usable results for the RIFS Vendor certified value is a ldquomade tordquo value as determined by volumetric andor gravimetric measurements of National Institute of Standards and Technology (NIST) Standard

5

de maximis inc

Reference Materials or NIST traceable reference material andor analytically verified by the vendor Vendor Performance Acceptance Limits are based on actual historical data collected in the vendors proficiency testing program The limits reflect any inherent biases in the methods used to establish the limits and closely approximate a 95 confidence interval of the performance that experienced laboratories should achieve using accepted environmental methods

An additional statistical evaluation of the analysis of uranium by ICPMS and alpha spectroscopy was performed using linear regression Linear regression mathematically approximates the relationship between two variables and provides an equation to predict a dependent variable from an independent variable Since ICPMS was selected as the primary analytical method for uranium is was assigned as the independent variable (x) and alpha spectroscopy was assigned as the dependent variable (y)

Furthermore the relationship between these two methods was anticipated to be linear such that y = x ideally so the y-intercept was set to equal zero for linear regression During the Fall 2004 RI phase of the investigation 172 soil and sediment samples were collected and analyzed by both analytical methods Of these samples 3 were determined to be from the super-depleted subset by ICPMS and were excluded from linear regression as the results of these samples are questionable

The results of linear regression for this data set are an estimated model of y (alpha spectroscopy) = 10604 x (ICPMS) with an R2 value of 09049 R2 measures the strength of the linear regression model with a value of 1 indicating a perfect linear fit so the results indicate a very strong fit In addition the 95 confidence interval for the slope ranges from 10080 to 11128 which indicates that linear regression modeled alpha spectroscopy results are generally 6 higher than ICPMS results with a range of 1 to 11 (95 confidence) The results of the linear regression analysis are shown in the attached Figure 2 (note that the scale is logarithmic in order to allow better display of all the data points across the range of results) From Figure 2 it is evident that there is substantial scatter in the low concentration range (less than 10 mgkg) and if these data points are removed from the linear regression evaluation there is no significant change in the results since the y-intercept was set to equal zero

These linear regression modeling results provide strong evidence that when the entire population of uranium analysis results for both ICPMS and alpha spectroscopy are considered there is good agreement between the two methods The relative difference is not significant when compared to the expected accuracy of the analytical techniques (+25)

In response to the comment regarding use of alpha spectroscopy values rather than ICPMS values to make remedial decisions it is clear that the commenter is not sufficiently familiar with the CERCLA investigation and remediation process The purpose of the remedial investigation (RI) is to collect data to evaluate the potential risks to human health and the environment associated with chemicals of potential concerns (COPCs) at the Site The data from the two methods show good overall correlation but as indicated there is a discrepancy if only the 1-50 ppm U-238 results are considered

6

de maximis inc

It is important to understand how the RI data will be evaluated and interpreted moving forward in the CERCLA process with respect to the concern that some areas might be missed for remediation The RI data set will be used to derive human exposure point concentrations (EPCs) for COPCs within the various media and exposure areas at the Site EPCs will be derived in accordance with EPA guidance which generally use 95 upper confidence level concentrations for soil analytes For uranium this will be performed using the ICPMS data set As a component of the risk assessment uncertainties associated with the analytical data are evaluated evaluation of alpha spectroscopy outliers as well as evaluation of the effect of uranium enrichment (eg ranging from 02 to 07) on risk estimates will be part of this process

If based on this process a given exposure areamedia has risks that EPA deems to be unacceptable (generally gt 10-4 cancer risk or non-cancer hazard index gt1 based on the cumulative risk from all COPCs in a given mediumexposure area) then preliminary remediation goals (PRGs) will be developed for the COPCs that contribute most substantially to the total risk In turn the PRGs will then be used to evaluate areas and volumes for which remedial alternatives will be considered in the Feasibility Study (FS) Areas for remediation will consider a variety of factors when determining the overall dimensions and they typically will be drawn to include RI data points that are equal to or greater than the PRGs

EPA will then select final remediation goals in the Record of Decision (ROD) The Remedial DesignRemedial Action (RDRA) process occurs after the ROD During the RD areas and volumes for remediation will be revisited based on the final remediation goals typically resulting in identification of a minimum target areavolume that will be remediated As part of the RD process a Demonstration of Compliance Plan will be developed This document sets forth how the contractor will demonstrate that all targeted media has been successfully remediated For a site like NMI we would expect that different instrumentation analytical methods may be used on Site to evaluate the effectiveness of remediation during the RA (eg gamma spectroscopy or XRF which provides faster results than ICPMS or alpha spectroscopy) followed by confirmatory sampling sent to an off-site laboratory Correlation of these instrument results to laboratory analytical methods will be presented at the appropriate time Given this process (CERCLA) and the multiple reviews by EPA State of Massachusetts agencies and public groups it is very unlikely that any areas with unacceptable risk levels will be missed for remediation

With respect to uranium background concentrations the total uranium values calculated using alpha spectroscopy values is correct in the CREW Technical Note (2006) however this background value is based on a theoretical calculation While this calculation is correct de maximis inc (and EPA via MampE split samples) have shown total uranium (reflected in the background evaluation) directly measured by ICPMS is lower due to inherent differences between the two methods and the fact that ICPMS is measuring ldquoenvironmentally availablerdquo concentrations Furthermore the background concentration at the Site is not likely to play a significant role since risk-based remediation levels greater than background are expected to be established for Site clean-up

7

de maximis inc

Lastly the intent for referencing the urine study was to show an independent evaluation that also concludes ICPMS has lower sensitivity for uranium and a greater ability to distinguish between natural depleted and enriched uranium Further it supports the argument that some variability between ICPMS and alpha spectroscopy analysis is not unexpected

In conclusion the ICPMS analysis provides lower results as compared to alpha spectroscopy analysis results however as discussed in this response (and described in detail in MampE uranium evaluation memo) neither method is in error

bull Background uranium concentrations determined from actual ICPMS measurements show environmentally available U background is approximately 13 ppm (EPA MampE-split samples confirm this)

bull lsquoSuper-depletedrsquo points are anomalies and these samples will be recollected and reanalyzed however remaining total U concentrations are usable and ICPMS results generated for the RIFS are reliable

bull Urine analysis referenced in the Technical Memorandum is applicable in that it supports the use of ICPMS for evaluating presence of U235 and calculating enrichments (for future samples once alpha spectroscopy analysis is discontinued)

bull Alpha spectroscopy analysis analyses for the Site were intended to verify enrichment to ensure that no enriched uranium was missed by ICPMS and to confirm signature of uranium and thorium at the site This has been done therefore future alpha spectroscopy analysis is not needed

Response to Section 3 ndash Using enrichment rather than total U for determination of contamination

Using total uranium by mass rather than enrichment to define nature and extent of contamination is being performed at this Site because the determination of enrichment at total uranium concentrations in the background range (eg 075 ndash 25 mgkg) is uncertain particularly for alpha spectroscopy analysis methods where the total propagated uncertainty of measurements is nearly equal to the actual instrument reading This is clearly demonstrated in the scatter observed in data plots provided by CREW (2006) for total uranium in this range de maximis inc concurs that at 5 mgkg a large percentage of the mass should be depleted uranium and for the purposes of defining the nature and extent of contamination based solely on isotopic composition a value of 5 would be inappropriate However the RI is using a value of 21 mgkg as a threshold for the definition of the nature and extent of contamination The nature and extent of uranium like all substances at the Site is being defined by risk-based levels (background levels are only used when background values are higher than risk-based levels) Therefore for total uranium concentrations less than the RI screening level the

8

de maximis inc

source of the uranium (whether depleted or natural uranium) is not relevant For risk assessment purposes all uranium will be considered and the contribution to risk associated with background constituents of concern will be considered after the total risk is calculated

Measurements of environmental data rarely have a normal (Gaussian) distribution and typically are lognormal distributions (Gilbert 1987) EPA guidance indicates that ldquoit is unusual to encounter environmental data sets that are normally distributedrdquo (EPA 2002) and ldquoenvironmental data commonly exhibit frequency distributions that are non-negative and skewed with heavy or long right tailsrdquo (EPA 2000) The composition of uranium (natural or depleted) will be used during risk assessment but it has limited usefulness otherwise However the isotopic composition of uranium will be one of the issues considered during data evaluation to avoid the sort of issues pointed out in the CREW Technical Note

Response to Section 4 - Thorium evaluation

de maximis inc re-iterates that the Technical Memorandum shows thorium-232 and thorium-228 concentrations are roughly equal (and considered within the analytical variability for alpha spectroscopy) and that thorium-230 at the Site is not related to Site activities Furthermore uranium-232 (from recycled uranium) is not present at the site as it is primarily a concern with enriched uranium Even if present it would not be in significant enough quantities to impact thorium-228 concentrations

Refer to previous discussion pertaining to ICPMS versus alpha spectroscopy analysis information Also note that ICPMS instrument sensitivity for thorium is very high due to the high mass to charge ratio (ie it is a good responder) Naturally occurring thorium is 100 Th-232 No data quality issues were observed during data validation and data has been deemed usable

Lastly thorium was used at the Site but at significantly lower amounts when compared to uranium Therefore low concentrations (within background concentration range) of thorium are not unexpected and historical data have demonstrated this

Response to Section 5 - Other radionuclides

There is no way to know the exact pedigree of the depleted uranium processed at the Site or to determine which portion of the depleted uranium might have come from reprocessed fuel rods and therefore contain small amounts of reactor generated radionuclides Studies of depleted uranium have shown that byproduct radionuclides in uranium are present at very low levels In addition dose assessments of these trace contaminants have shown that the radiological dose from these radionuclides is insignificant compared to that from uranium These conclusions were documented in a January 2000 Department of the Army analysis of transuranics and other contaminants in depleted uranium armor (Attachment 3) Of particular relevance is the fact that the Department of the Army analysis took samples of DU billets from Starmet for this study The Army concluded that ldquoThe levels of transuranics elements (TRUs) and fission

9

de maximis inc

products (FP) are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is saferdquo If the TRUs and FP contribute less than a one percent increase in dose for DU metal then they would contribute an even more negligible dose at the 10s to 1000s of mgkg of DU present in NMI Site media With respect to the HTDR analysis during the RI historical data indicated that the holding basin and former drain outfall regions of the bog were the most likely areas to have significant concentrations of uranium therefore the RI work plan identified locations to be sampled in these areas (FSP 2004) Further the approach outlined in the RI was to evaluate for HTDR in samples with elevated concentrations of uranium not necessarily the highest concentrations There were other AOIs where uranium was detected in elevated concentrations but no sample was collected for HTDR However since the holding basin was in use for the vast majority of time that uranium was actively processed at the site and no HTDR were detected in the six samples from this area it is unlikely that HTDR would be identified in other areas of the site

Samples for HTDRs were collected according to the RI (FSP 2004) such that 3 of the samples from AOI 1 were above the water table and 3 of the samples were from below the water table one sample was collected from each of the AOI 6 former drain outfall areas of the bog and one sample collected from the buried drum excavation (AOI 2) These HTDR samples are representative for the areas investigated (AOI 1 AOI 2 and AOI 6) and of the site based on the distribution of total uranium results described below

bull Total uranium in the HTDR sample locations from AOI 1 ranged from 38 to 511 mgkg while total uranium for all samples in AOI 1 ranged from 11 to 672 mgkg Only 25 samples from all samples in AOI 1 had total uranium concentrations greater than 100 mgkg and 4 of the 6 HTDR samples were from this subset

bull Total uranium in the HTDR sample locations from AOI 6 were 25 mgkg at former drain outfall 1 and 4550 mgkg at drain outfall 2 while total uranium for all samples in AOI 6 ranged from 16 to 4550 mgkg Only 10 samples from all samples in AOI 6 had total uranium concentrations greater than 100 mgkg and one HTDR sample was from the maximum total uranium sample location

bull The HTDR sample from AOI 2 drum excavation bottom had a total uranium concentration of 18 mgkg while total uranium for all excavation sides and bottom ranged from 14 to 49 mgkg

Only a few samples have uranium concentrations greater than 1000 mgkg and 6 of the 8 are from samples deemed as lsquosuper-depletedrsquo These 6 locations will be re-sampled as part of Phase 1C since the ICPMS analysis results are considered an anomaly Sporadic low detections of HTDRs do not necessarily indicate the presence of HTDRs as false positive results are common for these analyses Although some HTDRs have been detected in the past they are sporadic (not all samples have detections) and do not indicate the presence of HTDRs at significant levels

10

de maximis inc

It is also important to note that the historical data provided in the RI (PSOP 2005) was used to assist in design of the RI (FSP 2004) Data were compiled from many sources of information and represent samples collected analyzed and reported using various sample collection procedures laboratory analytical methods and data reporting conventions Data compilation activities involved both manual hand-entries and electronic entries when applicable Based on a review of the electronic database severe data quality limitations were noted including unknown analytical quality and validation levels missing location coordinates inconsistent media codes unknown detection limits and radionuclide uncertainty measurements inconsistent andor incorrect reporting of radionuclides using mass-based and activity-based units These limitations were considered significant during the scoping of the RIFS Work Plan therefore the historical data were used primarily to evaluate the spatial distribution of parameters and magnitude of detected concentrations The limitations of the historical data were used to provide the basis for identifying data gaps and scoping the RIFS As such the FSP included analysis for HTDRs at the areas of the Site known to be impacted by historical practices

With respect to comments provided on U-232 and U-236 these uranium isotopes are produced during the fission of uranium (as are plutonium americium and the rest of the HTDRs) Since none of these other fission products have been detected at the Site it is not credible that either uranium-232 or uranium-236 would be present

Response to Section 6 - Radium

The historical radium in groundwater data is suspect for several reasons including the previous discussion on historical data provided in the RI (PSOP 2005) Groundwater samples were collected from 13 wells at the Site (single sampling event) using bailers as part of this investigation All samples were screened for radioactivity by gross alpha beta analysis Only one sample exceeded the action level (gt 10 pCil gross alpha) with results of 50 pCil gross alpha and 55 pCil gross beta and it was analyzed for isotopic uranium Ra-226 and Ra-228 The results were 74 pCil U-234 025 pCil U-235 037 pCil U-238 12 pCil Ra-226 and 7 pCil Ra-228 These values are from Table 10 of the report and it appears that the Ra-226 value was incorrectly transcribed in the CREW Technical Note as 37 pCil The isotopic uranium results do not match those expected as U-234 is elevated and results for U-235 and U-238 are almost equal Therefore the analytical results for this sample are questionable Furthermore all the rest of the ground water samples had gross alpha results of 3 pCil or less indicating that there are no significant amounts of radium in groundwater

In addition a limited literature review was performed to assess the potential for leaching of radium from subsurface solids by nitrate at the NMI Site The literature demonstrates that radium may be found in a handful of aqueous phase complexes The predominant species of radium in aqueous solution is the divalent cation Ra2+ but other complexes such as RaOH+ RaCl+ RaCO3 and RaSO4 may also be found (Beneš et al 1982 Beneš 1984) Radium appears to be the least likely of all the alkali earth metals (beryllium magnesium calcium strontium barium and radium) to form aqueous phase

11

de maximis inc

complexes (Kirby and Salutsky 1964) As such we are unclear as to the origin of Dr Landreacutersquos statement that ldquonitrates are known to influence greatly radium leaching from the soil into the groundwaterrdquo (Landreacute 2006) The salt Ra(NO3)2 is highly soluble (245 g100 g H2O Kirby and Salutsky 1964) but we have not found evidence to suggest that the species Ra(NO3)+ or Ra(NO3)2 exist in aqueous solution Lauria et al (2004) present radionuclide and water chemistry data from groundwater samples taken from the Buena Lagoon in Brazil A plot of Ra-226 and Ra-228 concentrations versus nitrate concentration is presented in Figure 1 These data demonstrate that both Ra-226 and Ra-228 are almost completely uncorrelated with nitrate at this site as evidenced by correlation coefficients (r2) of approximately 005 and 004 respectively Szabo et al (1997) also report that no correlation exists between nitrate and Ra-226 in 17 groundwater samples taken from the Kirkwood-Cohansey aquifer system in New Jersey although Ra-228 appears to be weakly correlated with nitrate in the same data set (r2 = 018)

Figure 1 Correlation between aqueous radium and nitrate (Lauria et al 2004)

r 2 = 005

r 2 = 004

1E-10 1E-09 1E-08 1E-07 1E-06 1E-05 1E-04

0 5 10 15 20 25

Nitrate (mgL)

Ra

(microg

L)

Ra-226 Ra-228 Linear (Ra-226) Linear (Ra-228)

The mobility of radium in the subsurface appears to be controlled primarily by sorption and precipitation (Lauria et al 2004) High values of aqueous phase radium have been observed in groundwater samples with either increased salinity or reduced pH (Kraemer and Reed 1984 Lauria and Godoy 2002) Demir (1989) demonstrated the increased mobility of radium in acidic environments due to the reduced sorptive capacity of sediments at low pH Figure 2 is a plot of radium concentration versus sample pH from the same data set (Lauria et al 2004)

12

de maximis inc

Figure 2 Correlation between aqueous radium and pH (Lauria et al 2004) 226Ra = -4E-06pH + 3E-05

r2 = 058

228Ra = -5E-08pH + 4E-07 r2 = 032

1E-10 1E-09 1E-08 1E-07 1E-06 1E-05 1E-04

3 4 5 6 7

pH

Ra

(microg

L)

Ra-226 Ra-228 Linear (Ra-226) Linear (Ra-228)

Here a stronger correlation between Ra-226 concentration and pH exists (r2 = 058) along with an observable trend of decreasing aqueous phase radium with increasing pH Szabo et al (1997) also report a negative correlation (r2 = 030) between Ra-228 and pH but no significant correlation between Ra-226 and pH Divalent cations such as calcium and magnesium compete with radium for a limited number of surface complexation and adsorption sites (Szabo et al 1997) As lime (a mixture potentially containing calcium oxide calcium hydroxide and calcium carbonate) was co-disposed with uranium and nitric acid at the Holding Basin nitrate may serve as a site-specific tracer for conditions that may lead to the release of adsorbed or surface-complexed radium from the solid phase although no comparable calcium ldquoplumerdquo has been defined unlike nitrate in groundwater Therefore we suggest that Dr Landreacutersquos statement could be qualified by stating that low pH values are known to influence the leaching of radium from aquifer solids However as discussed above we do not see a compelling reason to add radium to the analytical list at this time

Please contact me if you have any questions

Sincerely

Bruce Thompson

cc Respondents Bennie Underwood PE de maximis inc inc Nelson Walter PE MACTEC Inc Peter Zeeb PhD PG GeoSyntec Consultants Inc

13

de maximis inc

Figure 1 ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

Figure 2 Comparison of Alpha Spectroscopy and ICP-MS Uranium Analysis

Attachment 1 Metcalf amp Eddy Memorandum ndash Uranium Analysis comparison of alpha spectrometry and ICP-MS results

Attachment 2 QAPP SOPs for Alpha Spectroscopy and ICPMS Sample Preparation

Attachment 3 Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor Department of the Army January 19 2000

References

Beneš P Migration of radium in the terrestrial hydrosphere The behavior of radium in waterways and aquifers Technical Reports Series No 310 International Atomic Energy Agency IAEA-TECDOC-301 Vienna Austrai pp 119-173 1984

Beneš P M Obdržaacutelek M Čejchanovaacute The physicochemical forms of tracers of radium in aqueous solutions containing chlorides sulfates and carbonates Radiochem Radioanal Letters 50(4) pp 227-242 1982

CREW 2006 ldquoComments on Radiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Technical Note No 2006-01 January 6 2006

Demir I Temporal variations and sources of Pb Co Cu Ni Fe and Mo in shallow ground water of the McDonalds Branch basin Lebanon State Forest Burlington County New Jersey New Jersey Geological Survey Report 20 19 p 1989

EPA 2000 Guidance for Data Quality Assessment Practical Methods for Data Analysis EPA QAG-9 July 2000

EPA 2002 Guidance for Comparing Background and Chemical Concentrations in Soil for CERCLA Site EPA 540-R-01-003 September 2002

FSP 2004 Field Sampling Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Gilbert 1989 Statistical Methods for Environmental Pollution Monitoring Richard O Gilbert 1987

Kirby H W M L Salutsky The Radiochemistry of Radium National Academy of Sciences ndash National Research Council NAS-NS 3057 205 p 1964

14

de maximis inc

Kramer T F and D F Reid The occurrence and behavior of radium in saline formation water of the US Gulf Coast Region Isotope Geoscience 2 pp 153shy174 1984

Landreacute V Comments on ldquoRadiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Nuclear Metals Superfund Site CREW Technical Note 2006-01 January 6 2006

Lauria D C and J M O Godoy Abnormal high natural radium concentration in surface waters Journal of Environmental Radioactivity 61(2) pp 27-37 2002

Lauria D C R M R Almeida O Sracek Behavior of radium thorium and uranium in groundwater near the Buena Lagoon in the Coastal Zone of the State of Rio de Janeiro Brazil Environmental Geology 47 pp 11-19 200

MACTEC 2005 Radiological Review of Fall 2004 RI Analytical Data Nuclear Metals Superfund Site Concord Massachusetts September 26 2004

MampE 2005 Memorandum from Andrew Schkuta to Ed Conroy Subject Uranium Analysis comparison of alpha spectrometry and ICPMS results December 15 2005

PSOP Project Summary and Operations Plan Nuclear Metals Superfund Site Concord Massachusetts April 15 2005

QAPP 2004 Quality Assurance Project Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Szabo Z D E Rice C L MacLeod and T H Barringer Relation of distribution of radium nitrate and pesticides to agricultural land use and depth Kirkwood-Cohansey aquifer system New Jersey Coastal Plain 1990-1991 US Geological Survey Water-Resources Investigations Report 96-4165A 119 p 1997

15

bull

bull bull

bull bull

bull bull bull

bull bull

bull bull

bull

500

1500

Figure 1

Con

cent

ratio

n (m

gkg

)

ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

1000

0

0 5 10 15 20 25 30 35 40

Samples

Figure 2

Alp

ha S

pect

osco

py (m

gkg

)

Fall 2004 Total Uranium No Super-Depleted

1000

100

10

1

y = 10604x R2 = 09049

1 10 100 1000

ICPMS (mgkg)

Fall 2004 Super-Depleted Linear Regression

Attachment 1Connolly Jayme

From Taylor Mel issagepam ail epa gov Sent Wednesday December 21 2005 932AM To brucetdemaximiscom jhuntdemaximiscom Subject Fw Uranium Analysis Issues

Rad-versus-ICPMS RADvsiCPMS_soilse MEvsPRP_ICPMSP memodoc (158 dPDF (29 KB) OF (25 KB)

brucejohn here is mampes take on the rns vs alpha spec analysis discrepancy

Forwarded by MelissaG TaylorR1USEPAUS on 12212005 0930 AM

Conroy Ed ltEdConroyregrn-ea ecomcomgt To

MelissaG TaylorR1USEPAUSregEPA 12152005 0416 cc PM

Subject FW Uranium Analysis Issues

Melissa

Heres Andys take on the alpha spec vs ICPMS debate along with a couple of graphs showing correlation between the two methods and MampE split data vs PRP data

Please call me on the cell phone tomorrow to discuss

Ed

Edward A Conroy PE

Project Manager

METCALF amp EDDYJAECOM

7812246880 ph

6172832942 cell

1

ed conroym-e aecom com Attachment 1

-----Original Message----shyFrom Schkuta Andrew Sent Thursday December 15 2005 358 PM To Conroy Ed Subject Uranium Analysis Issues

Ed

Please find attached a memorandum that summarizes my review of why ICP-MS and alpha spec data may not necessarily agree

ltltRad-versus-ICPMS memodocgtgt

In addition I have attached 2 preliminary charts that have the following draft information

ltltRADvsiCPMS soilsedPDFgtgt

The RADvsiCPMS soilsed chart is a draft plot of MampEs alpha spec versus ICP-MS results when data was available for both samples It shows the same general trend as the PRP data with fairly good agreement between the two analyses There appears to be a slight positive bias for alpha spec data at low concentrations but this is expected as discussed in my memo

ltltMEvsPRP ICPMSPDFgtgt

The MEvsPRP ICPMS is a draft plot of MampE vs the PRPs total uranium results as measured by ICP-MS -With the exception of a few outliers which will be further investigated there appears to be good agreement between MampEs and the PRPs data As a note the median percent difference between MampEs and the PRPs data was 1

Andrew Schkuta CHMM

Technical Specialist - Chemistry

METCALF amp EDDYJAECOM

701 Edgewater Drive

Wakefield MA 01880

(781) 224-6353

(781) 245-6293 fax

2

(See attached file Rad-versus-ICPMS Attachment 1memodoc) (See attached file RADvsiCPMS soilsedPDF) (See attached file MEvsPRP ICPMSPDF)

3

Attachment 1

Metcalf amp Eddy Inc

METCALFampEDDY

Memorandum

Date December 15 2005

To Ed Conroy

From Andrew Schkuta

Subject Uranium Analysis comparison of alpha spectrometry and ICP-MS results

File WA151 Nuclear Metals Distribution

Summary This memorandum discusses the differences between uranium analysis conducted by two different methods alpha spectroscopy (alpha spec) and inductively coupled plasma shymass spectrometry (ICP-MS) In summary the alpha spec data is expected to be biased high compared to the ICP-MS data but this is not indicative that either method is being performed incorrectly

In the September 25 2005 technical memorandum Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts (MACTEC 2005) data was presented comparing alpha spec and ICP-MS analysis for uranium analysis and it was proposed that future uranium analysis be performed by ICP-MS only For samples analyzed that were collected through Spring 2005 all samples were analyzed by ICP-MS with a subset (20) analyzed by alpha spec It was noted that there appeared to be a positive bias in the alpha spec data compared to the ICP-MS data and some concern has been voiced if analysis by ICP-MS only would meet project objectives This memorandum provides background on why the two analytical methods may yield differing results and discusses how they may impact project quality objectives

There are two primary reasons why alpha spec is expected to yield results that may be positively biased

1 Sample Preparation The sample preparation method for alpha spec is more rigorous The preparation method in alpha spec breaks down the silica mineral structure of the sample By design the ICP-MS sample preparation does not break down the mineral structure and is intended to remove all metals that could become environmentally available

2 Sample Quanitation The results for alpha spec are adjusted for analyte loss by correcting the result based on the amount of tracer recovered (yield) ICP-MS data is absolute ie data is not corrected for loss during sample preparation

2 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

These two reasons are discussed in further detail as follows

Sample Preparation

Radionuclide analysis traditionally uses a sample preparation method that includes hydrofluoric acid (HF) to break down the sample so that all of the uranium present in the sample is put into solution On the other hand standard EPA methods for metals analysis including ICP-MS do not measure the total metals present in a sample For example method SW846-3050 (EPA 1996) (as specified by the PRP in their approved QAPP for metals analysis by ICP-MS) states

This method is not a total digestion technique for most samples It is a very strong acid digestion that will dissolve almost all elements that could become environmentally available By design elements bound in silicate structures are not normally dissolved by this procedure as they are not usually mobile in the environment If absolute total digestion is required use Method 3052

As noted by the method a SW846 method is available to do a total digestion (equivalent to the alpha spec method) as method 3052 However method 3052 (EPA 1996) states (emphasis added)

Note This technique is not appropriate for regulatory applications that require the use of leachate preparations (ie Method 3050 Method 3051 Method 1311 Method 1312 Method 1310 Method 1320 Method 1330 Method 3031 Method 3040) This method is appropriate for those applications requiring a total decomposition for research purposes (ie~ geological studiesr mass balancesr analysis of Standard Reference Materials) or in response to a regulation that requires total sample decomposition

Even though the ICP-MS preparation method does not yield total results it still may conservatively measure the amount of metals that are environmentally available In the Research Project Summary Ambient Levels of Metals in New Jersey Soils prepared by the New Jersey Department of Environmental Protection (NJDEP 2003) where samples were prepared using SW846-3050 procedures it states

Metal concentrations reported in this summary were acid-extractable metals not necessarily total metal concentrations The extraction method (USEPA Method 3050) is a vigorous extraction method designed to remove all metals that could become environmentally available In practice the extraction method likely overestimates metals that could become available since the sample is refluxed with both concentrated nitric acid and hydrogen peroxide However by design the method will not extract chemicals incorporated in silica minerals as they are usually not mobile in the environment Thus concentrations reported in these studies may be lower than those indicated from analyses using methods designed to measure total metal concentrations such as x-ray fluorescence methods

Although references to uranium were not located it is expected that uranium will have some of the same type of mobility characteristics as other metals in the soil matrix The paper Ground Water Issue Behavior of Metals in Soil published by EPAs Superfund Technology Support Center for Ground Water (EPA 1992) describes how metals are distributed within soil The paper describes how metals are partitioned in various pools and how a rigorous extraction using HF is required to release metals present in the structure of primary or secondary minerals It is

3 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

expected that native metals (ie background concentrations) can be present in these structures The paper indicates that for four metals examined (cadmium copper zinc and nickel) the amount of metals that was bound in the mineral structure ranges from approximately 15-60 It would be expected that naturally occurring uranium would have similar characteristics

The National Institutes of Standards amp Technology (NIST) recognizes the differences between total digestion and method SW846-3050 and has issued Addendum to Certificates for the Certificate of Analysis (NIST 2003) for some Standard Reference Materials (SRMs) The Addendum to Certificates for SRM 2711 a Montana agricultural soil indicates that trace metals were measured at approximately 40 to 90 of their certified value when analyzed using SW846-3050 as the preparation method Even though uranium is not listed as a target analyte it is reasonable to expect that is would behave in a similar fashion

Sample Quanitation

The preparation method for alpha spec analysis includes addition of a tracer (U-232) to sample and is carried through the entire analytical scheme The amount of tracer is measured during analysis and the results for target compounds (the other uranium isotopes) are corrected for the amount of tracer recovery (yield) For example if amount of tracer measured is 50 of what was initially added to the sample the sample results are all corrected by dividing by 050 In terms of method quality assurancequality control (QAQC) the PRP QAPP indicates that tracer recovery in the range of 15-125 is considered acceptable and the analytical system is considered in control In other words up to 85 of the compound can be lost during sample preparation and analysis and the data is considered meeting all QAQC objectives This method of analysis is the equivalent to the technique of isotope dilution which is used for some non-radiological environmental analyses (eg EPA Method 1625 Semivolatile Organic Compounds by Isotope Dilution GCMS) Analyses using this technique although considered the most accurate are not widely employed due to the complexity and expense of using the isotope dilution technique

ICP-MS analysis measures the concentration of analyte with no correction for potential losses However QAQC procedures are in place to make sure the losses are not considered significant In addition to laboratory control samples matrix spike analysis measures the recovery for the target analyte that was over-spiked into the sample The PRP QAPP states that matrix spike recovery should be within the range of 75-125 Therefore up to 25 of the analyte could be lost and the method has still met the QAQC objectives If recovery is less than this then validation procedures acknowledge that the data may be biased low and sample results are flagged as estimated

Therefore because the alpha spec data is corrected for analyte loss during the analytical process while ICP-MS data is not some amount of positive bias is expected to be present in alpha spec data when compared to ICP-MS data

Considerations

In evaluating which analysis is appropriate for samples from the project neither method is inherently better than the other The following factors should be considered

1 What is being measured ICP-MS measures mass directly and data can be mathematically converted to activity Alpha spec measures activity directly and can be mathematically converted to mass

4 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

2 What are the objectives of the data in terms of measuring total versus environmentally available The ICP-MS method specified in the PRP QAPP measures uranium concentrations in the same manner that EPA has measured other metals concentrations Should uranium mass be measured in a typical way that will be used for measuring other metals of concern at the site (eg copper beryllium arsenic etc)

3 If ICP-MS is more cost effective that alpha spec should the option of using preparation method SW846-3052 be considered This should yield total uranium data similar to alpha spec analysis while still using the efficiencies of ICP-MS analysis

4 What units will the cleanup level for the site be in If its in mass units then ICPshyMS analysis may be more appropriate If its in activity units then alpha spec analysis may be warranted

References

Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts Technical Memorandum from Heath Downey to Bruce Thompson de maxim is inc September 26 2005 (MACTEC 2005)

Test Methods for Evaluating Solid Waste PhysicalChemical Methods SW-846 Third Edition with updates EPA 1996

Ambient Levels of Metals in New Jersey Soils Environmental Assessment and Risk Analysis Element Research Project Summary Division of Science Research and Technology State of New Jersey Department of Environmental Protection May2003

Ground Water Issue Behavior of Metals in Soil EPA Superfund Technology Support Center for Ground Water EPA540S-92018 October 1993

Certificate of Analysis Standard Reference Material 2711 Montana Soil National Institute of Standards amp Technology July 18 2003

bull bull

bull bull bullbull bull

10000

MampE SoilSediment Results Total Uranium mgKg

bull

bull - ~-~-----~~---~~-~--~~1000

en aE

I I bull Total a u

I -

bull bull 100 1shy -middot- -middot bullbull bull

Ybull bull

10 100 1000 10000

Alpha Spec

Attachm

ent 1

10 -

100000

TOTAL URANIUM BY ICP-MS mgKg

bull

bull 10000

I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 2: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

de maximis inc

bull The fall 2004 radionuclide data supports the conclusion that future alpha spectroscopy analysis is not needed as the isotopic signature for both uranium and thorium has been defined and the lack of enriched uranium was demonstrated The purpose for the 5 (1 alpha spectroscopy20 ICPMS) alpha spectroscopy split analysis was to verify the original assumptions made in RIFS work plan (PSOP 2005) with respect to uranium and thorium isotopic profiles verify the uranium profile resembles natural or depleted isotopic mix and determine the thorium profile in site media so that the contribution of thorium can be distinguished from the background profile and that associated with uranium decay

bull Uranium and thorium results for ICPMS tend to be lower as compared to alpha spectroscopy analysis however this is not unexpected as ICPMS results are representative of environmental conditions at the site whereas alpha spectroscopy analysis results are representative of environmental and natural conditions combined

bull ICPMS uranium results for a small sub-set of samples are an analytical anomaly however this fact does not cast doubt on the entire ICPMS data set

bull The variability in results does not cast doubt on the reliability of the laboratory results

bull The nature and extent of uranium at the Site as well as clean-up levels will be based on the mass of uranium present not enrichment

bull Samples for HTDR analysis were collected as specified in the RIFS Work Plan (FSP 2004) and several of these samples were collected in areas with elevated uranium concentrations Further testing for these radionuclides is not needed

bull Uranium-232 and Uranium-236 measurements are not needed as HTDR analysis demonstrated the lack of reactor-produced radionuclides

bull Radium measurements are not needed as there was no historical use at the site and the historical result appears to be an anomaly

Response to Section 2 - Soil and sediment analysis by ICPMS versus alpha spectroscopy

The lsquosuper-depletedrsquo uranium points shown in Figure 1 of the CREW Technical Note (CREW 2006) are an anomaly in the ICPMS analysis and lsquosuper-depletedrsquo uranium does not exist However given the large amount of ICPMS uranium data points collected from the site (651 uranium data points for soil and sediment from fall 2004 data set) it is reasonable and not unexpected to find a few outliers in every sampling event Upon further investigation it was determined the analytical results for these data points are from two (2) laboratory sample delivery groups (SDGs) Although results for

2

de maximis inc

these samples have undergone data validation procedures no analytical or transcription errors were observed Furthermore even though anomalies are identified results from all other samples (including other samples within the same SDGs) are valid and usable These limited anomalies should not cast doubt on all other uranium data generated using ICPMS

A detailed review was performed on the uranium results for the lsquosuper-depletedrsquo samples The review identified the following however no specific reason for the anomaly was identified

bull Sample results are limited to only 2 SDGs for ICPMS

bull U-235 and U-238 isotopes were reported from the same analytical run

bull Calculated enrichment values (ratio U-235U-238) from ICPMS instrument printshyouts (expressed as ugL) matched calculated enrichment values from mass-converted results (expressed as mgkg)

bull No transcription errors were found sample calculations and dilutions were verified as compared to Report of Analysis Forms

bull Calibration curve slopes for each isotope were very similar (U-235 = 000007084 and U-238 = 000007368)

bull Laboratory preparation and batching worksheets showed that although samples were prepped batched and analyzed at similar times no distinct pattern was recognized that could lead to an explanation for errors (ie other samples were prepped batched and analyzed with the lsquosuper-depletedrsquo samples and results for these samples showed either depleted or natural isotopic profiles)

bull ICPMS uranium concentrations for these two SDGs were plotted (See Figure 1) in ascending order Note the discrete inflection point for sample concentrations starting at approximately 100 mgkg These points (plus 2 additional points greater than 1600 mgkg) correspond to the samples termed lsquosuper-depletedrsquo

To better assess the uranium concentrations at the ldquosuper-depletedrsquo sampling locations the lsquosuper-depletedrsquo samples will be re-collected as part of the Phase IC soil investigation

As demonstrated in the Technical Memorandum and as supported by an independent evaluation of uranium split-sample results prepared by Metcalf and Eddy (MampE) on behalf of EPA (refer to Attachment 1) the alpha spectroscopy analysis results generally provide higher results as compared to ICPMS analysis As discussed below and reshyiterated in the MampE evaluation the sample preparation and digestion techniques for the two methods are different as is sample quantitation

3

de maximis inc

The ICPMS digestion procedure is designed to evaluate metals that could become ldquoenvironmentallyrdquo available whereas the alpha spectroscopy digestion is an extremely aggressive technique where complete mineral dissolution is performed As a result the generally higher uranium concentrations reported for the alpha spectroscopy analyses are not unexpected and may be explained by the liberation of geologically-bound uranium during digestion

The following information describes the methodologies and differences between the two analytical methods

bull Preparation homogenization and digestion ndash laboratory SOPs for the RIFS (QAPP 2004) are presented in Attachment 2 for these techniques The alpha spectroscopy preparation technique (SOP GL-RAD-015) was adapted from the DOE HASL Manual and uses a larger aliquot of sample and dries grinds ashes the sample prior to acid digestion The alpha spectroscopy analysis uses complete dissolution (HNO3 and HF) The ICPMS technique (SOP GL-MA-Eshy009) follows the EPA digestion method approved for the RIFS (QAPP 2004) and includes homogenization of the wet sample (smaller aliquot) and digestion using hot acid (HNO3 and H2O2)

bull Sample Quantitation - alpha spectroscopy method quantitation procedure includes adjustment of instrument results based on tracer R (percent yield is factored into reported result) ICPMS method quantitation procedure does not include applying correction factors for any bias observed during analysis (ie an absolute value is reported)

bull The ICPMS hot acid digestion technique used for the RIFS is a standard EPA digestion technique used at hundreds of CERCLA sites and is a widely accepted and aggressive method that is appropriate for a wide range of metals including uranium and thorium Furthermore as documented in the laboratory ICPMS SOP (GL-MA-E-014) uranium isotopes and thorium are suitable metals for analysis

bull As shown in Attachment 1 there is a very strong comparison between uranium split-sample results reported by de maximis inc and MampE which demonstrates the representativeness and usability of ICPMS data for the Site

Another factor to consider as part of the comparison of analytical methods is the resolution of uranium by each method The primary isotope of natural and depleted uranium is U-238 which has a low specific activity (ie it takes a large amount of mass to increase the amount of radioactivity) For example each increase in radioactivity by 1 pCig of U-238 results in an increase in mass of 3 mgkg Due to the fact that ICPMS has better resolution and sensitivity at lower mass concentrations it was chosen (and approved by EPA) as the quantitative analytical method for uranium for the RIFS

4

de maximis inc

EPA metals digestion procedures for all analyses (whether ICP ICPMS cold vapor) do not ldquounder-leachrdquo samples rather digestion procedures are designed to evaluate ldquoenvironmentally availablerdquo concentrations As such uranium and thorium would be analyzed as for other potential metal contaminants at the site (Cu Be Pb Hg etc)

While it is in fact possible that some disposed uranium could resist the hot nitric digestion it would also certainly resist leaching by rainwater infiltration or groundwater flushing and be less biologically available if ingested

To assess laboratory performance of the QAPP-approved methods for the analysis of uranium and thorium for the project performance evaluation (PE) samples have been submitted to the laboratory As part of the Phase IA and Phase 1B investigation programs blind PE samples (quality control standard that is of a composition and concentration not known to the laboratory) were submitted to the laboratory for analysis Results of the analyses and comparison to vendor spike concentrations are provided below

Investigation Analysis Parameter Laboratory Result Certified Value (Vendor Spike)

QC PALs

Phase 1A ICPMS Uranium (total)

32 26 24 mgkg1 22 mgkg NA

Phase 1A ICPMS Thorium 229 159 162 110 mgkg1

131 mgkg NA

Phase 1A Alpha Spec

Uraniumshy238

313 pCig 339 pCig NA

Phase 1A Alpha Spec

Uraniumshy234

34 pCig 339 pCig NA

Phase 1A Alpha Spec

Uraniumshy235

0252 U pCig2 016 pCig NA

Phase 1A Alpha Spec

Thorium-230 261 pCig 134 pCig NA

Phase 1B Alpha Spec

Uranium (total)

1184 pCiL 117 pCiL 852 ndash 138 pCiL

Phase 1B ICPMS Uranium 95 mgkg 126 mgkg 834 ndash 144 mgkg

Phase 1B ICPMS Uranium 606 ugL 635 ugL 525 ndash 745 ugL

1 = results provided for multiple runs performed by the laboratory 2 = not detected laboratory minimum detectable activity is greater than amount present in PE sample NA = not available

As seen in this table the laboratory provided acceptable results (ie results generally showed good agreement to vendor certified values and results were reported within QC Performance Acceptance Limits of the vendor result (when applicable) as determined by peer laboratory measurements) indicating the QAPP methods used for uranium and thorium analysis are capable of providing usable results for the RIFS Vendor certified value is a ldquomade tordquo value as determined by volumetric andor gravimetric measurements of National Institute of Standards and Technology (NIST) Standard

5

de maximis inc

Reference Materials or NIST traceable reference material andor analytically verified by the vendor Vendor Performance Acceptance Limits are based on actual historical data collected in the vendors proficiency testing program The limits reflect any inherent biases in the methods used to establish the limits and closely approximate a 95 confidence interval of the performance that experienced laboratories should achieve using accepted environmental methods

An additional statistical evaluation of the analysis of uranium by ICPMS and alpha spectroscopy was performed using linear regression Linear regression mathematically approximates the relationship between two variables and provides an equation to predict a dependent variable from an independent variable Since ICPMS was selected as the primary analytical method for uranium is was assigned as the independent variable (x) and alpha spectroscopy was assigned as the dependent variable (y)

Furthermore the relationship between these two methods was anticipated to be linear such that y = x ideally so the y-intercept was set to equal zero for linear regression During the Fall 2004 RI phase of the investigation 172 soil and sediment samples were collected and analyzed by both analytical methods Of these samples 3 were determined to be from the super-depleted subset by ICPMS and were excluded from linear regression as the results of these samples are questionable

The results of linear regression for this data set are an estimated model of y (alpha spectroscopy) = 10604 x (ICPMS) with an R2 value of 09049 R2 measures the strength of the linear regression model with a value of 1 indicating a perfect linear fit so the results indicate a very strong fit In addition the 95 confidence interval for the slope ranges from 10080 to 11128 which indicates that linear regression modeled alpha spectroscopy results are generally 6 higher than ICPMS results with a range of 1 to 11 (95 confidence) The results of the linear regression analysis are shown in the attached Figure 2 (note that the scale is logarithmic in order to allow better display of all the data points across the range of results) From Figure 2 it is evident that there is substantial scatter in the low concentration range (less than 10 mgkg) and if these data points are removed from the linear regression evaluation there is no significant change in the results since the y-intercept was set to equal zero

These linear regression modeling results provide strong evidence that when the entire population of uranium analysis results for both ICPMS and alpha spectroscopy are considered there is good agreement between the two methods The relative difference is not significant when compared to the expected accuracy of the analytical techniques (+25)

In response to the comment regarding use of alpha spectroscopy values rather than ICPMS values to make remedial decisions it is clear that the commenter is not sufficiently familiar with the CERCLA investigation and remediation process The purpose of the remedial investigation (RI) is to collect data to evaluate the potential risks to human health and the environment associated with chemicals of potential concerns (COPCs) at the Site The data from the two methods show good overall correlation but as indicated there is a discrepancy if only the 1-50 ppm U-238 results are considered

6

de maximis inc

It is important to understand how the RI data will be evaluated and interpreted moving forward in the CERCLA process with respect to the concern that some areas might be missed for remediation The RI data set will be used to derive human exposure point concentrations (EPCs) for COPCs within the various media and exposure areas at the Site EPCs will be derived in accordance with EPA guidance which generally use 95 upper confidence level concentrations for soil analytes For uranium this will be performed using the ICPMS data set As a component of the risk assessment uncertainties associated with the analytical data are evaluated evaluation of alpha spectroscopy outliers as well as evaluation of the effect of uranium enrichment (eg ranging from 02 to 07) on risk estimates will be part of this process

If based on this process a given exposure areamedia has risks that EPA deems to be unacceptable (generally gt 10-4 cancer risk or non-cancer hazard index gt1 based on the cumulative risk from all COPCs in a given mediumexposure area) then preliminary remediation goals (PRGs) will be developed for the COPCs that contribute most substantially to the total risk In turn the PRGs will then be used to evaluate areas and volumes for which remedial alternatives will be considered in the Feasibility Study (FS) Areas for remediation will consider a variety of factors when determining the overall dimensions and they typically will be drawn to include RI data points that are equal to or greater than the PRGs

EPA will then select final remediation goals in the Record of Decision (ROD) The Remedial DesignRemedial Action (RDRA) process occurs after the ROD During the RD areas and volumes for remediation will be revisited based on the final remediation goals typically resulting in identification of a minimum target areavolume that will be remediated As part of the RD process a Demonstration of Compliance Plan will be developed This document sets forth how the contractor will demonstrate that all targeted media has been successfully remediated For a site like NMI we would expect that different instrumentation analytical methods may be used on Site to evaluate the effectiveness of remediation during the RA (eg gamma spectroscopy or XRF which provides faster results than ICPMS or alpha spectroscopy) followed by confirmatory sampling sent to an off-site laboratory Correlation of these instrument results to laboratory analytical methods will be presented at the appropriate time Given this process (CERCLA) and the multiple reviews by EPA State of Massachusetts agencies and public groups it is very unlikely that any areas with unacceptable risk levels will be missed for remediation

With respect to uranium background concentrations the total uranium values calculated using alpha spectroscopy values is correct in the CREW Technical Note (2006) however this background value is based on a theoretical calculation While this calculation is correct de maximis inc (and EPA via MampE split samples) have shown total uranium (reflected in the background evaluation) directly measured by ICPMS is lower due to inherent differences between the two methods and the fact that ICPMS is measuring ldquoenvironmentally availablerdquo concentrations Furthermore the background concentration at the Site is not likely to play a significant role since risk-based remediation levels greater than background are expected to be established for Site clean-up

7

de maximis inc

Lastly the intent for referencing the urine study was to show an independent evaluation that also concludes ICPMS has lower sensitivity for uranium and a greater ability to distinguish between natural depleted and enriched uranium Further it supports the argument that some variability between ICPMS and alpha spectroscopy analysis is not unexpected

In conclusion the ICPMS analysis provides lower results as compared to alpha spectroscopy analysis results however as discussed in this response (and described in detail in MampE uranium evaluation memo) neither method is in error

bull Background uranium concentrations determined from actual ICPMS measurements show environmentally available U background is approximately 13 ppm (EPA MampE-split samples confirm this)

bull lsquoSuper-depletedrsquo points are anomalies and these samples will be recollected and reanalyzed however remaining total U concentrations are usable and ICPMS results generated for the RIFS are reliable

bull Urine analysis referenced in the Technical Memorandum is applicable in that it supports the use of ICPMS for evaluating presence of U235 and calculating enrichments (for future samples once alpha spectroscopy analysis is discontinued)

bull Alpha spectroscopy analysis analyses for the Site were intended to verify enrichment to ensure that no enriched uranium was missed by ICPMS and to confirm signature of uranium and thorium at the site This has been done therefore future alpha spectroscopy analysis is not needed

Response to Section 3 ndash Using enrichment rather than total U for determination of contamination

Using total uranium by mass rather than enrichment to define nature and extent of contamination is being performed at this Site because the determination of enrichment at total uranium concentrations in the background range (eg 075 ndash 25 mgkg) is uncertain particularly for alpha spectroscopy analysis methods where the total propagated uncertainty of measurements is nearly equal to the actual instrument reading This is clearly demonstrated in the scatter observed in data plots provided by CREW (2006) for total uranium in this range de maximis inc concurs that at 5 mgkg a large percentage of the mass should be depleted uranium and for the purposes of defining the nature and extent of contamination based solely on isotopic composition a value of 5 would be inappropriate However the RI is using a value of 21 mgkg as a threshold for the definition of the nature and extent of contamination The nature and extent of uranium like all substances at the Site is being defined by risk-based levels (background levels are only used when background values are higher than risk-based levels) Therefore for total uranium concentrations less than the RI screening level the

8

de maximis inc

source of the uranium (whether depleted or natural uranium) is not relevant For risk assessment purposes all uranium will be considered and the contribution to risk associated with background constituents of concern will be considered after the total risk is calculated

Measurements of environmental data rarely have a normal (Gaussian) distribution and typically are lognormal distributions (Gilbert 1987) EPA guidance indicates that ldquoit is unusual to encounter environmental data sets that are normally distributedrdquo (EPA 2002) and ldquoenvironmental data commonly exhibit frequency distributions that are non-negative and skewed with heavy or long right tailsrdquo (EPA 2000) The composition of uranium (natural or depleted) will be used during risk assessment but it has limited usefulness otherwise However the isotopic composition of uranium will be one of the issues considered during data evaluation to avoid the sort of issues pointed out in the CREW Technical Note

Response to Section 4 - Thorium evaluation

de maximis inc re-iterates that the Technical Memorandum shows thorium-232 and thorium-228 concentrations are roughly equal (and considered within the analytical variability for alpha spectroscopy) and that thorium-230 at the Site is not related to Site activities Furthermore uranium-232 (from recycled uranium) is not present at the site as it is primarily a concern with enriched uranium Even if present it would not be in significant enough quantities to impact thorium-228 concentrations

Refer to previous discussion pertaining to ICPMS versus alpha spectroscopy analysis information Also note that ICPMS instrument sensitivity for thorium is very high due to the high mass to charge ratio (ie it is a good responder) Naturally occurring thorium is 100 Th-232 No data quality issues were observed during data validation and data has been deemed usable

Lastly thorium was used at the Site but at significantly lower amounts when compared to uranium Therefore low concentrations (within background concentration range) of thorium are not unexpected and historical data have demonstrated this

Response to Section 5 - Other radionuclides

There is no way to know the exact pedigree of the depleted uranium processed at the Site or to determine which portion of the depleted uranium might have come from reprocessed fuel rods and therefore contain small amounts of reactor generated radionuclides Studies of depleted uranium have shown that byproduct radionuclides in uranium are present at very low levels In addition dose assessments of these trace contaminants have shown that the radiological dose from these radionuclides is insignificant compared to that from uranium These conclusions were documented in a January 2000 Department of the Army analysis of transuranics and other contaminants in depleted uranium armor (Attachment 3) Of particular relevance is the fact that the Department of the Army analysis took samples of DU billets from Starmet for this study The Army concluded that ldquoThe levels of transuranics elements (TRUs) and fission

9

de maximis inc

products (FP) are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is saferdquo If the TRUs and FP contribute less than a one percent increase in dose for DU metal then they would contribute an even more negligible dose at the 10s to 1000s of mgkg of DU present in NMI Site media With respect to the HTDR analysis during the RI historical data indicated that the holding basin and former drain outfall regions of the bog were the most likely areas to have significant concentrations of uranium therefore the RI work plan identified locations to be sampled in these areas (FSP 2004) Further the approach outlined in the RI was to evaluate for HTDR in samples with elevated concentrations of uranium not necessarily the highest concentrations There were other AOIs where uranium was detected in elevated concentrations but no sample was collected for HTDR However since the holding basin was in use for the vast majority of time that uranium was actively processed at the site and no HTDR were detected in the six samples from this area it is unlikely that HTDR would be identified in other areas of the site

Samples for HTDRs were collected according to the RI (FSP 2004) such that 3 of the samples from AOI 1 were above the water table and 3 of the samples were from below the water table one sample was collected from each of the AOI 6 former drain outfall areas of the bog and one sample collected from the buried drum excavation (AOI 2) These HTDR samples are representative for the areas investigated (AOI 1 AOI 2 and AOI 6) and of the site based on the distribution of total uranium results described below

bull Total uranium in the HTDR sample locations from AOI 1 ranged from 38 to 511 mgkg while total uranium for all samples in AOI 1 ranged from 11 to 672 mgkg Only 25 samples from all samples in AOI 1 had total uranium concentrations greater than 100 mgkg and 4 of the 6 HTDR samples were from this subset

bull Total uranium in the HTDR sample locations from AOI 6 were 25 mgkg at former drain outfall 1 and 4550 mgkg at drain outfall 2 while total uranium for all samples in AOI 6 ranged from 16 to 4550 mgkg Only 10 samples from all samples in AOI 6 had total uranium concentrations greater than 100 mgkg and one HTDR sample was from the maximum total uranium sample location

bull The HTDR sample from AOI 2 drum excavation bottom had a total uranium concentration of 18 mgkg while total uranium for all excavation sides and bottom ranged from 14 to 49 mgkg

Only a few samples have uranium concentrations greater than 1000 mgkg and 6 of the 8 are from samples deemed as lsquosuper-depletedrsquo These 6 locations will be re-sampled as part of Phase 1C since the ICPMS analysis results are considered an anomaly Sporadic low detections of HTDRs do not necessarily indicate the presence of HTDRs as false positive results are common for these analyses Although some HTDRs have been detected in the past they are sporadic (not all samples have detections) and do not indicate the presence of HTDRs at significant levels

10

de maximis inc

It is also important to note that the historical data provided in the RI (PSOP 2005) was used to assist in design of the RI (FSP 2004) Data were compiled from many sources of information and represent samples collected analyzed and reported using various sample collection procedures laboratory analytical methods and data reporting conventions Data compilation activities involved both manual hand-entries and electronic entries when applicable Based on a review of the electronic database severe data quality limitations were noted including unknown analytical quality and validation levels missing location coordinates inconsistent media codes unknown detection limits and radionuclide uncertainty measurements inconsistent andor incorrect reporting of radionuclides using mass-based and activity-based units These limitations were considered significant during the scoping of the RIFS Work Plan therefore the historical data were used primarily to evaluate the spatial distribution of parameters and magnitude of detected concentrations The limitations of the historical data were used to provide the basis for identifying data gaps and scoping the RIFS As such the FSP included analysis for HTDRs at the areas of the Site known to be impacted by historical practices

With respect to comments provided on U-232 and U-236 these uranium isotopes are produced during the fission of uranium (as are plutonium americium and the rest of the HTDRs) Since none of these other fission products have been detected at the Site it is not credible that either uranium-232 or uranium-236 would be present

Response to Section 6 - Radium

The historical radium in groundwater data is suspect for several reasons including the previous discussion on historical data provided in the RI (PSOP 2005) Groundwater samples were collected from 13 wells at the Site (single sampling event) using bailers as part of this investigation All samples were screened for radioactivity by gross alpha beta analysis Only one sample exceeded the action level (gt 10 pCil gross alpha) with results of 50 pCil gross alpha and 55 pCil gross beta and it was analyzed for isotopic uranium Ra-226 and Ra-228 The results were 74 pCil U-234 025 pCil U-235 037 pCil U-238 12 pCil Ra-226 and 7 pCil Ra-228 These values are from Table 10 of the report and it appears that the Ra-226 value was incorrectly transcribed in the CREW Technical Note as 37 pCil The isotopic uranium results do not match those expected as U-234 is elevated and results for U-235 and U-238 are almost equal Therefore the analytical results for this sample are questionable Furthermore all the rest of the ground water samples had gross alpha results of 3 pCil or less indicating that there are no significant amounts of radium in groundwater

In addition a limited literature review was performed to assess the potential for leaching of radium from subsurface solids by nitrate at the NMI Site The literature demonstrates that radium may be found in a handful of aqueous phase complexes The predominant species of radium in aqueous solution is the divalent cation Ra2+ but other complexes such as RaOH+ RaCl+ RaCO3 and RaSO4 may also be found (Beneš et al 1982 Beneš 1984) Radium appears to be the least likely of all the alkali earth metals (beryllium magnesium calcium strontium barium and radium) to form aqueous phase

11

de maximis inc

complexes (Kirby and Salutsky 1964) As such we are unclear as to the origin of Dr Landreacutersquos statement that ldquonitrates are known to influence greatly radium leaching from the soil into the groundwaterrdquo (Landreacute 2006) The salt Ra(NO3)2 is highly soluble (245 g100 g H2O Kirby and Salutsky 1964) but we have not found evidence to suggest that the species Ra(NO3)+ or Ra(NO3)2 exist in aqueous solution Lauria et al (2004) present radionuclide and water chemistry data from groundwater samples taken from the Buena Lagoon in Brazil A plot of Ra-226 and Ra-228 concentrations versus nitrate concentration is presented in Figure 1 These data demonstrate that both Ra-226 and Ra-228 are almost completely uncorrelated with nitrate at this site as evidenced by correlation coefficients (r2) of approximately 005 and 004 respectively Szabo et al (1997) also report that no correlation exists between nitrate and Ra-226 in 17 groundwater samples taken from the Kirkwood-Cohansey aquifer system in New Jersey although Ra-228 appears to be weakly correlated with nitrate in the same data set (r2 = 018)

Figure 1 Correlation between aqueous radium and nitrate (Lauria et al 2004)

r 2 = 005

r 2 = 004

1E-10 1E-09 1E-08 1E-07 1E-06 1E-05 1E-04

0 5 10 15 20 25

Nitrate (mgL)

Ra

(microg

L)

Ra-226 Ra-228 Linear (Ra-226) Linear (Ra-228)

The mobility of radium in the subsurface appears to be controlled primarily by sorption and precipitation (Lauria et al 2004) High values of aqueous phase radium have been observed in groundwater samples with either increased salinity or reduced pH (Kraemer and Reed 1984 Lauria and Godoy 2002) Demir (1989) demonstrated the increased mobility of radium in acidic environments due to the reduced sorptive capacity of sediments at low pH Figure 2 is a plot of radium concentration versus sample pH from the same data set (Lauria et al 2004)

12

de maximis inc

Figure 2 Correlation between aqueous radium and pH (Lauria et al 2004) 226Ra = -4E-06pH + 3E-05

r2 = 058

228Ra = -5E-08pH + 4E-07 r2 = 032

1E-10 1E-09 1E-08 1E-07 1E-06 1E-05 1E-04

3 4 5 6 7

pH

Ra

(microg

L)

Ra-226 Ra-228 Linear (Ra-226) Linear (Ra-228)

Here a stronger correlation between Ra-226 concentration and pH exists (r2 = 058) along with an observable trend of decreasing aqueous phase radium with increasing pH Szabo et al (1997) also report a negative correlation (r2 = 030) between Ra-228 and pH but no significant correlation between Ra-226 and pH Divalent cations such as calcium and magnesium compete with radium for a limited number of surface complexation and adsorption sites (Szabo et al 1997) As lime (a mixture potentially containing calcium oxide calcium hydroxide and calcium carbonate) was co-disposed with uranium and nitric acid at the Holding Basin nitrate may serve as a site-specific tracer for conditions that may lead to the release of adsorbed or surface-complexed radium from the solid phase although no comparable calcium ldquoplumerdquo has been defined unlike nitrate in groundwater Therefore we suggest that Dr Landreacutersquos statement could be qualified by stating that low pH values are known to influence the leaching of radium from aquifer solids However as discussed above we do not see a compelling reason to add radium to the analytical list at this time

Please contact me if you have any questions

Sincerely

Bruce Thompson

cc Respondents Bennie Underwood PE de maximis inc inc Nelson Walter PE MACTEC Inc Peter Zeeb PhD PG GeoSyntec Consultants Inc

13

de maximis inc

Figure 1 ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

Figure 2 Comparison of Alpha Spectroscopy and ICP-MS Uranium Analysis

Attachment 1 Metcalf amp Eddy Memorandum ndash Uranium Analysis comparison of alpha spectrometry and ICP-MS results

Attachment 2 QAPP SOPs for Alpha Spectroscopy and ICPMS Sample Preparation

Attachment 3 Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor Department of the Army January 19 2000

References

Beneš P Migration of radium in the terrestrial hydrosphere The behavior of radium in waterways and aquifers Technical Reports Series No 310 International Atomic Energy Agency IAEA-TECDOC-301 Vienna Austrai pp 119-173 1984

Beneš P M Obdržaacutelek M Čejchanovaacute The physicochemical forms of tracers of radium in aqueous solutions containing chlorides sulfates and carbonates Radiochem Radioanal Letters 50(4) pp 227-242 1982

CREW 2006 ldquoComments on Radiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Technical Note No 2006-01 January 6 2006

Demir I Temporal variations and sources of Pb Co Cu Ni Fe and Mo in shallow ground water of the McDonalds Branch basin Lebanon State Forest Burlington County New Jersey New Jersey Geological Survey Report 20 19 p 1989

EPA 2000 Guidance for Data Quality Assessment Practical Methods for Data Analysis EPA QAG-9 July 2000

EPA 2002 Guidance for Comparing Background and Chemical Concentrations in Soil for CERCLA Site EPA 540-R-01-003 September 2002

FSP 2004 Field Sampling Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Gilbert 1989 Statistical Methods for Environmental Pollution Monitoring Richard O Gilbert 1987

Kirby H W M L Salutsky The Radiochemistry of Radium National Academy of Sciences ndash National Research Council NAS-NS 3057 205 p 1964

14

de maximis inc

Kramer T F and D F Reid The occurrence and behavior of radium in saline formation water of the US Gulf Coast Region Isotope Geoscience 2 pp 153shy174 1984

Landreacute V Comments on ldquoRadiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Nuclear Metals Superfund Site CREW Technical Note 2006-01 January 6 2006

Lauria D C and J M O Godoy Abnormal high natural radium concentration in surface waters Journal of Environmental Radioactivity 61(2) pp 27-37 2002

Lauria D C R M R Almeida O Sracek Behavior of radium thorium and uranium in groundwater near the Buena Lagoon in the Coastal Zone of the State of Rio de Janeiro Brazil Environmental Geology 47 pp 11-19 200

MACTEC 2005 Radiological Review of Fall 2004 RI Analytical Data Nuclear Metals Superfund Site Concord Massachusetts September 26 2004

MampE 2005 Memorandum from Andrew Schkuta to Ed Conroy Subject Uranium Analysis comparison of alpha spectrometry and ICPMS results December 15 2005

PSOP Project Summary and Operations Plan Nuclear Metals Superfund Site Concord Massachusetts April 15 2005

QAPP 2004 Quality Assurance Project Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Szabo Z D E Rice C L MacLeod and T H Barringer Relation of distribution of radium nitrate and pesticides to agricultural land use and depth Kirkwood-Cohansey aquifer system New Jersey Coastal Plain 1990-1991 US Geological Survey Water-Resources Investigations Report 96-4165A 119 p 1997

15

bull

bull bull

bull bull

bull bull bull

bull bull

bull bull

bull

500

1500

Figure 1

Con

cent

ratio

n (m

gkg

)

ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

1000

0

0 5 10 15 20 25 30 35 40

Samples

Figure 2

Alp

ha S

pect

osco

py (m

gkg

)

Fall 2004 Total Uranium No Super-Depleted

1000

100

10

1

y = 10604x R2 = 09049

1 10 100 1000

ICPMS (mgkg)

Fall 2004 Super-Depleted Linear Regression

Attachment 1Connolly Jayme

From Taylor Mel issagepam ail epa gov Sent Wednesday December 21 2005 932AM To brucetdemaximiscom jhuntdemaximiscom Subject Fw Uranium Analysis Issues

Rad-versus-ICPMS RADvsiCPMS_soilse MEvsPRP_ICPMSP memodoc (158 dPDF (29 KB) OF (25 KB)

brucejohn here is mampes take on the rns vs alpha spec analysis discrepancy

Forwarded by MelissaG TaylorR1USEPAUS on 12212005 0930 AM

Conroy Ed ltEdConroyregrn-ea ecomcomgt To

MelissaG TaylorR1USEPAUSregEPA 12152005 0416 cc PM

Subject FW Uranium Analysis Issues

Melissa

Heres Andys take on the alpha spec vs ICPMS debate along with a couple of graphs showing correlation between the two methods and MampE split data vs PRP data

Please call me on the cell phone tomorrow to discuss

Ed

Edward A Conroy PE

Project Manager

METCALF amp EDDYJAECOM

7812246880 ph

6172832942 cell

1

ed conroym-e aecom com Attachment 1

-----Original Message----shyFrom Schkuta Andrew Sent Thursday December 15 2005 358 PM To Conroy Ed Subject Uranium Analysis Issues

Ed

Please find attached a memorandum that summarizes my review of why ICP-MS and alpha spec data may not necessarily agree

ltltRad-versus-ICPMS memodocgtgt

In addition I have attached 2 preliminary charts that have the following draft information

ltltRADvsiCPMS soilsedPDFgtgt

The RADvsiCPMS soilsed chart is a draft plot of MampEs alpha spec versus ICP-MS results when data was available for both samples It shows the same general trend as the PRP data with fairly good agreement between the two analyses There appears to be a slight positive bias for alpha spec data at low concentrations but this is expected as discussed in my memo

ltltMEvsPRP ICPMSPDFgtgt

The MEvsPRP ICPMS is a draft plot of MampE vs the PRPs total uranium results as measured by ICP-MS -With the exception of a few outliers which will be further investigated there appears to be good agreement between MampEs and the PRPs data As a note the median percent difference between MampEs and the PRPs data was 1

Andrew Schkuta CHMM

Technical Specialist - Chemistry

METCALF amp EDDYJAECOM

701 Edgewater Drive

Wakefield MA 01880

(781) 224-6353

(781) 245-6293 fax

2

(See attached file Rad-versus-ICPMS Attachment 1memodoc) (See attached file RADvsiCPMS soilsedPDF) (See attached file MEvsPRP ICPMSPDF)

3

Attachment 1

Metcalf amp Eddy Inc

METCALFampEDDY

Memorandum

Date December 15 2005

To Ed Conroy

From Andrew Schkuta

Subject Uranium Analysis comparison of alpha spectrometry and ICP-MS results

File WA151 Nuclear Metals Distribution

Summary This memorandum discusses the differences between uranium analysis conducted by two different methods alpha spectroscopy (alpha spec) and inductively coupled plasma shymass spectrometry (ICP-MS) In summary the alpha spec data is expected to be biased high compared to the ICP-MS data but this is not indicative that either method is being performed incorrectly

In the September 25 2005 technical memorandum Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts (MACTEC 2005) data was presented comparing alpha spec and ICP-MS analysis for uranium analysis and it was proposed that future uranium analysis be performed by ICP-MS only For samples analyzed that were collected through Spring 2005 all samples were analyzed by ICP-MS with a subset (20) analyzed by alpha spec It was noted that there appeared to be a positive bias in the alpha spec data compared to the ICP-MS data and some concern has been voiced if analysis by ICP-MS only would meet project objectives This memorandum provides background on why the two analytical methods may yield differing results and discusses how they may impact project quality objectives

There are two primary reasons why alpha spec is expected to yield results that may be positively biased

1 Sample Preparation The sample preparation method for alpha spec is more rigorous The preparation method in alpha spec breaks down the silica mineral structure of the sample By design the ICP-MS sample preparation does not break down the mineral structure and is intended to remove all metals that could become environmentally available

2 Sample Quanitation The results for alpha spec are adjusted for analyte loss by correcting the result based on the amount of tracer recovered (yield) ICP-MS data is absolute ie data is not corrected for loss during sample preparation

2 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

These two reasons are discussed in further detail as follows

Sample Preparation

Radionuclide analysis traditionally uses a sample preparation method that includes hydrofluoric acid (HF) to break down the sample so that all of the uranium present in the sample is put into solution On the other hand standard EPA methods for metals analysis including ICP-MS do not measure the total metals present in a sample For example method SW846-3050 (EPA 1996) (as specified by the PRP in their approved QAPP for metals analysis by ICP-MS) states

This method is not a total digestion technique for most samples It is a very strong acid digestion that will dissolve almost all elements that could become environmentally available By design elements bound in silicate structures are not normally dissolved by this procedure as they are not usually mobile in the environment If absolute total digestion is required use Method 3052

As noted by the method a SW846 method is available to do a total digestion (equivalent to the alpha spec method) as method 3052 However method 3052 (EPA 1996) states (emphasis added)

Note This technique is not appropriate for regulatory applications that require the use of leachate preparations (ie Method 3050 Method 3051 Method 1311 Method 1312 Method 1310 Method 1320 Method 1330 Method 3031 Method 3040) This method is appropriate for those applications requiring a total decomposition for research purposes (ie~ geological studiesr mass balancesr analysis of Standard Reference Materials) or in response to a regulation that requires total sample decomposition

Even though the ICP-MS preparation method does not yield total results it still may conservatively measure the amount of metals that are environmentally available In the Research Project Summary Ambient Levels of Metals in New Jersey Soils prepared by the New Jersey Department of Environmental Protection (NJDEP 2003) where samples were prepared using SW846-3050 procedures it states

Metal concentrations reported in this summary were acid-extractable metals not necessarily total metal concentrations The extraction method (USEPA Method 3050) is a vigorous extraction method designed to remove all metals that could become environmentally available In practice the extraction method likely overestimates metals that could become available since the sample is refluxed with both concentrated nitric acid and hydrogen peroxide However by design the method will not extract chemicals incorporated in silica minerals as they are usually not mobile in the environment Thus concentrations reported in these studies may be lower than those indicated from analyses using methods designed to measure total metal concentrations such as x-ray fluorescence methods

Although references to uranium were not located it is expected that uranium will have some of the same type of mobility characteristics as other metals in the soil matrix The paper Ground Water Issue Behavior of Metals in Soil published by EPAs Superfund Technology Support Center for Ground Water (EPA 1992) describes how metals are distributed within soil The paper describes how metals are partitioned in various pools and how a rigorous extraction using HF is required to release metals present in the structure of primary or secondary minerals It is

3 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

expected that native metals (ie background concentrations) can be present in these structures The paper indicates that for four metals examined (cadmium copper zinc and nickel) the amount of metals that was bound in the mineral structure ranges from approximately 15-60 It would be expected that naturally occurring uranium would have similar characteristics

The National Institutes of Standards amp Technology (NIST) recognizes the differences between total digestion and method SW846-3050 and has issued Addendum to Certificates for the Certificate of Analysis (NIST 2003) for some Standard Reference Materials (SRMs) The Addendum to Certificates for SRM 2711 a Montana agricultural soil indicates that trace metals were measured at approximately 40 to 90 of their certified value when analyzed using SW846-3050 as the preparation method Even though uranium is not listed as a target analyte it is reasonable to expect that is would behave in a similar fashion

Sample Quanitation

The preparation method for alpha spec analysis includes addition of a tracer (U-232) to sample and is carried through the entire analytical scheme The amount of tracer is measured during analysis and the results for target compounds (the other uranium isotopes) are corrected for the amount of tracer recovery (yield) For example if amount of tracer measured is 50 of what was initially added to the sample the sample results are all corrected by dividing by 050 In terms of method quality assurancequality control (QAQC) the PRP QAPP indicates that tracer recovery in the range of 15-125 is considered acceptable and the analytical system is considered in control In other words up to 85 of the compound can be lost during sample preparation and analysis and the data is considered meeting all QAQC objectives This method of analysis is the equivalent to the technique of isotope dilution which is used for some non-radiological environmental analyses (eg EPA Method 1625 Semivolatile Organic Compounds by Isotope Dilution GCMS) Analyses using this technique although considered the most accurate are not widely employed due to the complexity and expense of using the isotope dilution technique

ICP-MS analysis measures the concentration of analyte with no correction for potential losses However QAQC procedures are in place to make sure the losses are not considered significant In addition to laboratory control samples matrix spike analysis measures the recovery for the target analyte that was over-spiked into the sample The PRP QAPP states that matrix spike recovery should be within the range of 75-125 Therefore up to 25 of the analyte could be lost and the method has still met the QAQC objectives If recovery is less than this then validation procedures acknowledge that the data may be biased low and sample results are flagged as estimated

Therefore because the alpha spec data is corrected for analyte loss during the analytical process while ICP-MS data is not some amount of positive bias is expected to be present in alpha spec data when compared to ICP-MS data

Considerations

In evaluating which analysis is appropriate for samples from the project neither method is inherently better than the other The following factors should be considered

1 What is being measured ICP-MS measures mass directly and data can be mathematically converted to activity Alpha spec measures activity directly and can be mathematically converted to mass

4 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

2 What are the objectives of the data in terms of measuring total versus environmentally available The ICP-MS method specified in the PRP QAPP measures uranium concentrations in the same manner that EPA has measured other metals concentrations Should uranium mass be measured in a typical way that will be used for measuring other metals of concern at the site (eg copper beryllium arsenic etc)

3 If ICP-MS is more cost effective that alpha spec should the option of using preparation method SW846-3052 be considered This should yield total uranium data similar to alpha spec analysis while still using the efficiencies of ICP-MS analysis

4 What units will the cleanup level for the site be in If its in mass units then ICPshyMS analysis may be more appropriate If its in activity units then alpha spec analysis may be warranted

References

Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts Technical Memorandum from Heath Downey to Bruce Thompson de maxim is inc September 26 2005 (MACTEC 2005)

Test Methods for Evaluating Solid Waste PhysicalChemical Methods SW-846 Third Edition with updates EPA 1996

Ambient Levels of Metals in New Jersey Soils Environmental Assessment and Risk Analysis Element Research Project Summary Division of Science Research and Technology State of New Jersey Department of Environmental Protection May2003

Ground Water Issue Behavior of Metals in Soil EPA Superfund Technology Support Center for Ground Water EPA540S-92018 October 1993

Certificate of Analysis Standard Reference Material 2711 Montana Soil National Institute of Standards amp Technology July 18 2003

bull bull

bull bull bullbull bull

10000

MampE SoilSediment Results Total Uranium mgKg

bull

bull - ~-~-----~~---~~-~--~~1000

en aE

I I bull Total a u

I -

bull bull 100 1shy -middot- -middot bullbull bull

Ybull bull

10 100 1000 10000

Alpha Spec

Attachm

ent 1

10 -

100000

TOTAL URANIUM BY ICP-MS mgKg

bull

bull 10000

I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 3: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

de maximis inc

these samples have undergone data validation procedures no analytical or transcription errors were observed Furthermore even though anomalies are identified results from all other samples (including other samples within the same SDGs) are valid and usable These limited anomalies should not cast doubt on all other uranium data generated using ICPMS

A detailed review was performed on the uranium results for the lsquosuper-depletedrsquo samples The review identified the following however no specific reason for the anomaly was identified

bull Sample results are limited to only 2 SDGs for ICPMS

bull U-235 and U-238 isotopes were reported from the same analytical run

bull Calculated enrichment values (ratio U-235U-238) from ICPMS instrument printshyouts (expressed as ugL) matched calculated enrichment values from mass-converted results (expressed as mgkg)

bull No transcription errors were found sample calculations and dilutions were verified as compared to Report of Analysis Forms

bull Calibration curve slopes for each isotope were very similar (U-235 = 000007084 and U-238 = 000007368)

bull Laboratory preparation and batching worksheets showed that although samples were prepped batched and analyzed at similar times no distinct pattern was recognized that could lead to an explanation for errors (ie other samples were prepped batched and analyzed with the lsquosuper-depletedrsquo samples and results for these samples showed either depleted or natural isotopic profiles)

bull ICPMS uranium concentrations for these two SDGs were plotted (See Figure 1) in ascending order Note the discrete inflection point for sample concentrations starting at approximately 100 mgkg These points (plus 2 additional points greater than 1600 mgkg) correspond to the samples termed lsquosuper-depletedrsquo

To better assess the uranium concentrations at the ldquosuper-depletedrsquo sampling locations the lsquosuper-depletedrsquo samples will be re-collected as part of the Phase IC soil investigation

As demonstrated in the Technical Memorandum and as supported by an independent evaluation of uranium split-sample results prepared by Metcalf and Eddy (MampE) on behalf of EPA (refer to Attachment 1) the alpha spectroscopy analysis results generally provide higher results as compared to ICPMS analysis As discussed below and reshyiterated in the MampE evaluation the sample preparation and digestion techniques for the two methods are different as is sample quantitation

3

de maximis inc

The ICPMS digestion procedure is designed to evaluate metals that could become ldquoenvironmentallyrdquo available whereas the alpha spectroscopy digestion is an extremely aggressive technique where complete mineral dissolution is performed As a result the generally higher uranium concentrations reported for the alpha spectroscopy analyses are not unexpected and may be explained by the liberation of geologically-bound uranium during digestion

The following information describes the methodologies and differences between the two analytical methods

bull Preparation homogenization and digestion ndash laboratory SOPs for the RIFS (QAPP 2004) are presented in Attachment 2 for these techniques The alpha spectroscopy preparation technique (SOP GL-RAD-015) was adapted from the DOE HASL Manual and uses a larger aliquot of sample and dries grinds ashes the sample prior to acid digestion The alpha spectroscopy analysis uses complete dissolution (HNO3 and HF) The ICPMS technique (SOP GL-MA-Eshy009) follows the EPA digestion method approved for the RIFS (QAPP 2004) and includes homogenization of the wet sample (smaller aliquot) and digestion using hot acid (HNO3 and H2O2)

bull Sample Quantitation - alpha spectroscopy method quantitation procedure includes adjustment of instrument results based on tracer R (percent yield is factored into reported result) ICPMS method quantitation procedure does not include applying correction factors for any bias observed during analysis (ie an absolute value is reported)

bull The ICPMS hot acid digestion technique used for the RIFS is a standard EPA digestion technique used at hundreds of CERCLA sites and is a widely accepted and aggressive method that is appropriate for a wide range of metals including uranium and thorium Furthermore as documented in the laboratory ICPMS SOP (GL-MA-E-014) uranium isotopes and thorium are suitable metals for analysis

bull As shown in Attachment 1 there is a very strong comparison between uranium split-sample results reported by de maximis inc and MampE which demonstrates the representativeness and usability of ICPMS data for the Site

Another factor to consider as part of the comparison of analytical methods is the resolution of uranium by each method The primary isotope of natural and depleted uranium is U-238 which has a low specific activity (ie it takes a large amount of mass to increase the amount of radioactivity) For example each increase in radioactivity by 1 pCig of U-238 results in an increase in mass of 3 mgkg Due to the fact that ICPMS has better resolution and sensitivity at lower mass concentrations it was chosen (and approved by EPA) as the quantitative analytical method for uranium for the RIFS

4

de maximis inc

EPA metals digestion procedures for all analyses (whether ICP ICPMS cold vapor) do not ldquounder-leachrdquo samples rather digestion procedures are designed to evaluate ldquoenvironmentally availablerdquo concentrations As such uranium and thorium would be analyzed as for other potential metal contaminants at the site (Cu Be Pb Hg etc)

While it is in fact possible that some disposed uranium could resist the hot nitric digestion it would also certainly resist leaching by rainwater infiltration or groundwater flushing and be less biologically available if ingested

To assess laboratory performance of the QAPP-approved methods for the analysis of uranium and thorium for the project performance evaluation (PE) samples have been submitted to the laboratory As part of the Phase IA and Phase 1B investigation programs blind PE samples (quality control standard that is of a composition and concentration not known to the laboratory) were submitted to the laboratory for analysis Results of the analyses and comparison to vendor spike concentrations are provided below

Investigation Analysis Parameter Laboratory Result Certified Value (Vendor Spike)

QC PALs

Phase 1A ICPMS Uranium (total)

32 26 24 mgkg1 22 mgkg NA

Phase 1A ICPMS Thorium 229 159 162 110 mgkg1

131 mgkg NA

Phase 1A Alpha Spec

Uraniumshy238

313 pCig 339 pCig NA

Phase 1A Alpha Spec

Uraniumshy234

34 pCig 339 pCig NA

Phase 1A Alpha Spec

Uraniumshy235

0252 U pCig2 016 pCig NA

Phase 1A Alpha Spec

Thorium-230 261 pCig 134 pCig NA

Phase 1B Alpha Spec

Uranium (total)

1184 pCiL 117 pCiL 852 ndash 138 pCiL

Phase 1B ICPMS Uranium 95 mgkg 126 mgkg 834 ndash 144 mgkg

Phase 1B ICPMS Uranium 606 ugL 635 ugL 525 ndash 745 ugL

1 = results provided for multiple runs performed by the laboratory 2 = not detected laboratory minimum detectable activity is greater than amount present in PE sample NA = not available

As seen in this table the laboratory provided acceptable results (ie results generally showed good agreement to vendor certified values and results were reported within QC Performance Acceptance Limits of the vendor result (when applicable) as determined by peer laboratory measurements) indicating the QAPP methods used for uranium and thorium analysis are capable of providing usable results for the RIFS Vendor certified value is a ldquomade tordquo value as determined by volumetric andor gravimetric measurements of National Institute of Standards and Technology (NIST) Standard

5

de maximis inc

Reference Materials or NIST traceable reference material andor analytically verified by the vendor Vendor Performance Acceptance Limits are based on actual historical data collected in the vendors proficiency testing program The limits reflect any inherent biases in the methods used to establish the limits and closely approximate a 95 confidence interval of the performance that experienced laboratories should achieve using accepted environmental methods

An additional statistical evaluation of the analysis of uranium by ICPMS and alpha spectroscopy was performed using linear regression Linear regression mathematically approximates the relationship between two variables and provides an equation to predict a dependent variable from an independent variable Since ICPMS was selected as the primary analytical method for uranium is was assigned as the independent variable (x) and alpha spectroscopy was assigned as the dependent variable (y)

Furthermore the relationship between these two methods was anticipated to be linear such that y = x ideally so the y-intercept was set to equal zero for linear regression During the Fall 2004 RI phase of the investigation 172 soil and sediment samples were collected and analyzed by both analytical methods Of these samples 3 were determined to be from the super-depleted subset by ICPMS and were excluded from linear regression as the results of these samples are questionable

The results of linear regression for this data set are an estimated model of y (alpha spectroscopy) = 10604 x (ICPMS) with an R2 value of 09049 R2 measures the strength of the linear regression model with a value of 1 indicating a perfect linear fit so the results indicate a very strong fit In addition the 95 confidence interval for the slope ranges from 10080 to 11128 which indicates that linear regression modeled alpha spectroscopy results are generally 6 higher than ICPMS results with a range of 1 to 11 (95 confidence) The results of the linear regression analysis are shown in the attached Figure 2 (note that the scale is logarithmic in order to allow better display of all the data points across the range of results) From Figure 2 it is evident that there is substantial scatter in the low concentration range (less than 10 mgkg) and if these data points are removed from the linear regression evaluation there is no significant change in the results since the y-intercept was set to equal zero

These linear regression modeling results provide strong evidence that when the entire population of uranium analysis results for both ICPMS and alpha spectroscopy are considered there is good agreement between the two methods The relative difference is not significant when compared to the expected accuracy of the analytical techniques (+25)

In response to the comment regarding use of alpha spectroscopy values rather than ICPMS values to make remedial decisions it is clear that the commenter is not sufficiently familiar with the CERCLA investigation and remediation process The purpose of the remedial investigation (RI) is to collect data to evaluate the potential risks to human health and the environment associated with chemicals of potential concerns (COPCs) at the Site The data from the two methods show good overall correlation but as indicated there is a discrepancy if only the 1-50 ppm U-238 results are considered

6

de maximis inc

It is important to understand how the RI data will be evaluated and interpreted moving forward in the CERCLA process with respect to the concern that some areas might be missed for remediation The RI data set will be used to derive human exposure point concentrations (EPCs) for COPCs within the various media and exposure areas at the Site EPCs will be derived in accordance with EPA guidance which generally use 95 upper confidence level concentrations for soil analytes For uranium this will be performed using the ICPMS data set As a component of the risk assessment uncertainties associated with the analytical data are evaluated evaluation of alpha spectroscopy outliers as well as evaluation of the effect of uranium enrichment (eg ranging from 02 to 07) on risk estimates will be part of this process

If based on this process a given exposure areamedia has risks that EPA deems to be unacceptable (generally gt 10-4 cancer risk or non-cancer hazard index gt1 based on the cumulative risk from all COPCs in a given mediumexposure area) then preliminary remediation goals (PRGs) will be developed for the COPCs that contribute most substantially to the total risk In turn the PRGs will then be used to evaluate areas and volumes for which remedial alternatives will be considered in the Feasibility Study (FS) Areas for remediation will consider a variety of factors when determining the overall dimensions and they typically will be drawn to include RI data points that are equal to or greater than the PRGs

EPA will then select final remediation goals in the Record of Decision (ROD) The Remedial DesignRemedial Action (RDRA) process occurs after the ROD During the RD areas and volumes for remediation will be revisited based on the final remediation goals typically resulting in identification of a minimum target areavolume that will be remediated As part of the RD process a Demonstration of Compliance Plan will be developed This document sets forth how the contractor will demonstrate that all targeted media has been successfully remediated For a site like NMI we would expect that different instrumentation analytical methods may be used on Site to evaluate the effectiveness of remediation during the RA (eg gamma spectroscopy or XRF which provides faster results than ICPMS or alpha spectroscopy) followed by confirmatory sampling sent to an off-site laboratory Correlation of these instrument results to laboratory analytical methods will be presented at the appropriate time Given this process (CERCLA) and the multiple reviews by EPA State of Massachusetts agencies and public groups it is very unlikely that any areas with unacceptable risk levels will be missed for remediation

With respect to uranium background concentrations the total uranium values calculated using alpha spectroscopy values is correct in the CREW Technical Note (2006) however this background value is based on a theoretical calculation While this calculation is correct de maximis inc (and EPA via MampE split samples) have shown total uranium (reflected in the background evaluation) directly measured by ICPMS is lower due to inherent differences between the two methods and the fact that ICPMS is measuring ldquoenvironmentally availablerdquo concentrations Furthermore the background concentration at the Site is not likely to play a significant role since risk-based remediation levels greater than background are expected to be established for Site clean-up

7

de maximis inc

Lastly the intent for referencing the urine study was to show an independent evaluation that also concludes ICPMS has lower sensitivity for uranium and a greater ability to distinguish between natural depleted and enriched uranium Further it supports the argument that some variability between ICPMS and alpha spectroscopy analysis is not unexpected

In conclusion the ICPMS analysis provides lower results as compared to alpha spectroscopy analysis results however as discussed in this response (and described in detail in MampE uranium evaluation memo) neither method is in error

bull Background uranium concentrations determined from actual ICPMS measurements show environmentally available U background is approximately 13 ppm (EPA MampE-split samples confirm this)

bull lsquoSuper-depletedrsquo points are anomalies and these samples will be recollected and reanalyzed however remaining total U concentrations are usable and ICPMS results generated for the RIFS are reliable

bull Urine analysis referenced in the Technical Memorandum is applicable in that it supports the use of ICPMS for evaluating presence of U235 and calculating enrichments (for future samples once alpha spectroscopy analysis is discontinued)

bull Alpha spectroscopy analysis analyses for the Site were intended to verify enrichment to ensure that no enriched uranium was missed by ICPMS and to confirm signature of uranium and thorium at the site This has been done therefore future alpha spectroscopy analysis is not needed

Response to Section 3 ndash Using enrichment rather than total U for determination of contamination

Using total uranium by mass rather than enrichment to define nature and extent of contamination is being performed at this Site because the determination of enrichment at total uranium concentrations in the background range (eg 075 ndash 25 mgkg) is uncertain particularly for alpha spectroscopy analysis methods where the total propagated uncertainty of measurements is nearly equal to the actual instrument reading This is clearly demonstrated in the scatter observed in data plots provided by CREW (2006) for total uranium in this range de maximis inc concurs that at 5 mgkg a large percentage of the mass should be depleted uranium and for the purposes of defining the nature and extent of contamination based solely on isotopic composition a value of 5 would be inappropriate However the RI is using a value of 21 mgkg as a threshold for the definition of the nature and extent of contamination The nature and extent of uranium like all substances at the Site is being defined by risk-based levels (background levels are only used when background values are higher than risk-based levels) Therefore for total uranium concentrations less than the RI screening level the

8

de maximis inc

source of the uranium (whether depleted or natural uranium) is not relevant For risk assessment purposes all uranium will be considered and the contribution to risk associated with background constituents of concern will be considered after the total risk is calculated

Measurements of environmental data rarely have a normal (Gaussian) distribution and typically are lognormal distributions (Gilbert 1987) EPA guidance indicates that ldquoit is unusual to encounter environmental data sets that are normally distributedrdquo (EPA 2002) and ldquoenvironmental data commonly exhibit frequency distributions that are non-negative and skewed with heavy or long right tailsrdquo (EPA 2000) The composition of uranium (natural or depleted) will be used during risk assessment but it has limited usefulness otherwise However the isotopic composition of uranium will be one of the issues considered during data evaluation to avoid the sort of issues pointed out in the CREW Technical Note

Response to Section 4 - Thorium evaluation

de maximis inc re-iterates that the Technical Memorandum shows thorium-232 and thorium-228 concentrations are roughly equal (and considered within the analytical variability for alpha spectroscopy) and that thorium-230 at the Site is not related to Site activities Furthermore uranium-232 (from recycled uranium) is not present at the site as it is primarily a concern with enriched uranium Even if present it would not be in significant enough quantities to impact thorium-228 concentrations

Refer to previous discussion pertaining to ICPMS versus alpha spectroscopy analysis information Also note that ICPMS instrument sensitivity for thorium is very high due to the high mass to charge ratio (ie it is a good responder) Naturally occurring thorium is 100 Th-232 No data quality issues were observed during data validation and data has been deemed usable

Lastly thorium was used at the Site but at significantly lower amounts when compared to uranium Therefore low concentrations (within background concentration range) of thorium are not unexpected and historical data have demonstrated this

Response to Section 5 - Other radionuclides

There is no way to know the exact pedigree of the depleted uranium processed at the Site or to determine which portion of the depleted uranium might have come from reprocessed fuel rods and therefore contain small amounts of reactor generated radionuclides Studies of depleted uranium have shown that byproduct radionuclides in uranium are present at very low levels In addition dose assessments of these trace contaminants have shown that the radiological dose from these radionuclides is insignificant compared to that from uranium These conclusions were documented in a January 2000 Department of the Army analysis of transuranics and other contaminants in depleted uranium armor (Attachment 3) Of particular relevance is the fact that the Department of the Army analysis took samples of DU billets from Starmet for this study The Army concluded that ldquoThe levels of transuranics elements (TRUs) and fission

9

de maximis inc

products (FP) are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is saferdquo If the TRUs and FP contribute less than a one percent increase in dose for DU metal then they would contribute an even more negligible dose at the 10s to 1000s of mgkg of DU present in NMI Site media With respect to the HTDR analysis during the RI historical data indicated that the holding basin and former drain outfall regions of the bog were the most likely areas to have significant concentrations of uranium therefore the RI work plan identified locations to be sampled in these areas (FSP 2004) Further the approach outlined in the RI was to evaluate for HTDR in samples with elevated concentrations of uranium not necessarily the highest concentrations There were other AOIs where uranium was detected in elevated concentrations but no sample was collected for HTDR However since the holding basin was in use for the vast majority of time that uranium was actively processed at the site and no HTDR were detected in the six samples from this area it is unlikely that HTDR would be identified in other areas of the site

Samples for HTDRs were collected according to the RI (FSP 2004) such that 3 of the samples from AOI 1 were above the water table and 3 of the samples were from below the water table one sample was collected from each of the AOI 6 former drain outfall areas of the bog and one sample collected from the buried drum excavation (AOI 2) These HTDR samples are representative for the areas investigated (AOI 1 AOI 2 and AOI 6) and of the site based on the distribution of total uranium results described below

bull Total uranium in the HTDR sample locations from AOI 1 ranged from 38 to 511 mgkg while total uranium for all samples in AOI 1 ranged from 11 to 672 mgkg Only 25 samples from all samples in AOI 1 had total uranium concentrations greater than 100 mgkg and 4 of the 6 HTDR samples were from this subset

bull Total uranium in the HTDR sample locations from AOI 6 were 25 mgkg at former drain outfall 1 and 4550 mgkg at drain outfall 2 while total uranium for all samples in AOI 6 ranged from 16 to 4550 mgkg Only 10 samples from all samples in AOI 6 had total uranium concentrations greater than 100 mgkg and one HTDR sample was from the maximum total uranium sample location

bull The HTDR sample from AOI 2 drum excavation bottom had a total uranium concentration of 18 mgkg while total uranium for all excavation sides and bottom ranged from 14 to 49 mgkg

Only a few samples have uranium concentrations greater than 1000 mgkg and 6 of the 8 are from samples deemed as lsquosuper-depletedrsquo These 6 locations will be re-sampled as part of Phase 1C since the ICPMS analysis results are considered an anomaly Sporadic low detections of HTDRs do not necessarily indicate the presence of HTDRs as false positive results are common for these analyses Although some HTDRs have been detected in the past they are sporadic (not all samples have detections) and do not indicate the presence of HTDRs at significant levels

10

de maximis inc

It is also important to note that the historical data provided in the RI (PSOP 2005) was used to assist in design of the RI (FSP 2004) Data were compiled from many sources of information and represent samples collected analyzed and reported using various sample collection procedures laboratory analytical methods and data reporting conventions Data compilation activities involved both manual hand-entries and electronic entries when applicable Based on a review of the electronic database severe data quality limitations were noted including unknown analytical quality and validation levels missing location coordinates inconsistent media codes unknown detection limits and radionuclide uncertainty measurements inconsistent andor incorrect reporting of radionuclides using mass-based and activity-based units These limitations were considered significant during the scoping of the RIFS Work Plan therefore the historical data were used primarily to evaluate the spatial distribution of parameters and magnitude of detected concentrations The limitations of the historical data were used to provide the basis for identifying data gaps and scoping the RIFS As such the FSP included analysis for HTDRs at the areas of the Site known to be impacted by historical practices

With respect to comments provided on U-232 and U-236 these uranium isotopes are produced during the fission of uranium (as are plutonium americium and the rest of the HTDRs) Since none of these other fission products have been detected at the Site it is not credible that either uranium-232 or uranium-236 would be present

Response to Section 6 - Radium

The historical radium in groundwater data is suspect for several reasons including the previous discussion on historical data provided in the RI (PSOP 2005) Groundwater samples were collected from 13 wells at the Site (single sampling event) using bailers as part of this investigation All samples were screened for radioactivity by gross alpha beta analysis Only one sample exceeded the action level (gt 10 pCil gross alpha) with results of 50 pCil gross alpha and 55 pCil gross beta and it was analyzed for isotopic uranium Ra-226 and Ra-228 The results were 74 pCil U-234 025 pCil U-235 037 pCil U-238 12 pCil Ra-226 and 7 pCil Ra-228 These values are from Table 10 of the report and it appears that the Ra-226 value was incorrectly transcribed in the CREW Technical Note as 37 pCil The isotopic uranium results do not match those expected as U-234 is elevated and results for U-235 and U-238 are almost equal Therefore the analytical results for this sample are questionable Furthermore all the rest of the ground water samples had gross alpha results of 3 pCil or less indicating that there are no significant amounts of radium in groundwater

In addition a limited literature review was performed to assess the potential for leaching of radium from subsurface solids by nitrate at the NMI Site The literature demonstrates that radium may be found in a handful of aqueous phase complexes The predominant species of radium in aqueous solution is the divalent cation Ra2+ but other complexes such as RaOH+ RaCl+ RaCO3 and RaSO4 may also be found (Beneš et al 1982 Beneš 1984) Radium appears to be the least likely of all the alkali earth metals (beryllium magnesium calcium strontium barium and radium) to form aqueous phase

11

de maximis inc

complexes (Kirby and Salutsky 1964) As such we are unclear as to the origin of Dr Landreacutersquos statement that ldquonitrates are known to influence greatly radium leaching from the soil into the groundwaterrdquo (Landreacute 2006) The salt Ra(NO3)2 is highly soluble (245 g100 g H2O Kirby and Salutsky 1964) but we have not found evidence to suggest that the species Ra(NO3)+ or Ra(NO3)2 exist in aqueous solution Lauria et al (2004) present radionuclide and water chemistry data from groundwater samples taken from the Buena Lagoon in Brazil A plot of Ra-226 and Ra-228 concentrations versus nitrate concentration is presented in Figure 1 These data demonstrate that both Ra-226 and Ra-228 are almost completely uncorrelated with nitrate at this site as evidenced by correlation coefficients (r2) of approximately 005 and 004 respectively Szabo et al (1997) also report that no correlation exists between nitrate and Ra-226 in 17 groundwater samples taken from the Kirkwood-Cohansey aquifer system in New Jersey although Ra-228 appears to be weakly correlated with nitrate in the same data set (r2 = 018)

Figure 1 Correlation between aqueous radium and nitrate (Lauria et al 2004)

r 2 = 005

r 2 = 004

1E-10 1E-09 1E-08 1E-07 1E-06 1E-05 1E-04

0 5 10 15 20 25

Nitrate (mgL)

Ra

(microg

L)

Ra-226 Ra-228 Linear (Ra-226) Linear (Ra-228)

The mobility of radium in the subsurface appears to be controlled primarily by sorption and precipitation (Lauria et al 2004) High values of aqueous phase radium have been observed in groundwater samples with either increased salinity or reduced pH (Kraemer and Reed 1984 Lauria and Godoy 2002) Demir (1989) demonstrated the increased mobility of radium in acidic environments due to the reduced sorptive capacity of sediments at low pH Figure 2 is a plot of radium concentration versus sample pH from the same data set (Lauria et al 2004)

12

de maximis inc

Figure 2 Correlation between aqueous radium and pH (Lauria et al 2004) 226Ra = -4E-06pH + 3E-05

r2 = 058

228Ra = -5E-08pH + 4E-07 r2 = 032

1E-10 1E-09 1E-08 1E-07 1E-06 1E-05 1E-04

3 4 5 6 7

pH

Ra

(microg

L)

Ra-226 Ra-228 Linear (Ra-226) Linear (Ra-228)

Here a stronger correlation between Ra-226 concentration and pH exists (r2 = 058) along with an observable trend of decreasing aqueous phase radium with increasing pH Szabo et al (1997) also report a negative correlation (r2 = 030) between Ra-228 and pH but no significant correlation between Ra-226 and pH Divalent cations such as calcium and magnesium compete with radium for a limited number of surface complexation and adsorption sites (Szabo et al 1997) As lime (a mixture potentially containing calcium oxide calcium hydroxide and calcium carbonate) was co-disposed with uranium and nitric acid at the Holding Basin nitrate may serve as a site-specific tracer for conditions that may lead to the release of adsorbed or surface-complexed radium from the solid phase although no comparable calcium ldquoplumerdquo has been defined unlike nitrate in groundwater Therefore we suggest that Dr Landreacutersquos statement could be qualified by stating that low pH values are known to influence the leaching of radium from aquifer solids However as discussed above we do not see a compelling reason to add radium to the analytical list at this time

Please contact me if you have any questions

Sincerely

Bruce Thompson

cc Respondents Bennie Underwood PE de maximis inc inc Nelson Walter PE MACTEC Inc Peter Zeeb PhD PG GeoSyntec Consultants Inc

13

de maximis inc

Figure 1 ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

Figure 2 Comparison of Alpha Spectroscopy and ICP-MS Uranium Analysis

Attachment 1 Metcalf amp Eddy Memorandum ndash Uranium Analysis comparison of alpha spectrometry and ICP-MS results

Attachment 2 QAPP SOPs for Alpha Spectroscopy and ICPMS Sample Preparation

Attachment 3 Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor Department of the Army January 19 2000

References

Beneš P Migration of radium in the terrestrial hydrosphere The behavior of radium in waterways and aquifers Technical Reports Series No 310 International Atomic Energy Agency IAEA-TECDOC-301 Vienna Austrai pp 119-173 1984

Beneš P M Obdržaacutelek M Čejchanovaacute The physicochemical forms of tracers of radium in aqueous solutions containing chlorides sulfates and carbonates Radiochem Radioanal Letters 50(4) pp 227-242 1982

CREW 2006 ldquoComments on Radiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Technical Note No 2006-01 January 6 2006

Demir I Temporal variations and sources of Pb Co Cu Ni Fe and Mo in shallow ground water of the McDonalds Branch basin Lebanon State Forest Burlington County New Jersey New Jersey Geological Survey Report 20 19 p 1989

EPA 2000 Guidance for Data Quality Assessment Practical Methods for Data Analysis EPA QAG-9 July 2000

EPA 2002 Guidance for Comparing Background and Chemical Concentrations in Soil for CERCLA Site EPA 540-R-01-003 September 2002

FSP 2004 Field Sampling Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Gilbert 1989 Statistical Methods for Environmental Pollution Monitoring Richard O Gilbert 1987

Kirby H W M L Salutsky The Radiochemistry of Radium National Academy of Sciences ndash National Research Council NAS-NS 3057 205 p 1964

14

de maximis inc

Kramer T F and D F Reid The occurrence and behavior of radium in saline formation water of the US Gulf Coast Region Isotope Geoscience 2 pp 153shy174 1984

Landreacute V Comments on ldquoRadiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Nuclear Metals Superfund Site CREW Technical Note 2006-01 January 6 2006

Lauria D C and J M O Godoy Abnormal high natural radium concentration in surface waters Journal of Environmental Radioactivity 61(2) pp 27-37 2002

Lauria D C R M R Almeida O Sracek Behavior of radium thorium and uranium in groundwater near the Buena Lagoon in the Coastal Zone of the State of Rio de Janeiro Brazil Environmental Geology 47 pp 11-19 200

MACTEC 2005 Radiological Review of Fall 2004 RI Analytical Data Nuclear Metals Superfund Site Concord Massachusetts September 26 2004

MampE 2005 Memorandum from Andrew Schkuta to Ed Conroy Subject Uranium Analysis comparison of alpha spectrometry and ICPMS results December 15 2005

PSOP Project Summary and Operations Plan Nuclear Metals Superfund Site Concord Massachusetts April 15 2005

QAPP 2004 Quality Assurance Project Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Szabo Z D E Rice C L MacLeod and T H Barringer Relation of distribution of radium nitrate and pesticides to agricultural land use and depth Kirkwood-Cohansey aquifer system New Jersey Coastal Plain 1990-1991 US Geological Survey Water-Resources Investigations Report 96-4165A 119 p 1997

15

bull

bull bull

bull bull

bull bull bull

bull bull

bull bull

bull

500

1500

Figure 1

Con

cent

ratio

n (m

gkg

)

ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

1000

0

0 5 10 15 20 25 30 35 40

Samples

Figure 2

Alp

ha S

pect

osco

py (m

gkg

)

Fall 2004 Total Uranium No Super-Depleted

1000

100

10

1

y = 10604x R2 = 09049

1 10 100 1000

ICPMS (mgkg)

Fall 2004 Super-Depleted Linear Regression

Attachment 1Connolly Jayme

From Taylor Mel issagepam ail epa gov Sent Wednesday December 21 2005 932AM To brucetdemaximiscom jhuntdemaximiscom Subject Fw Uranium Analysis Issues

Rad-versus-ICPMS RADvsiCPMS_soilse MEvsPRP_ICPMSP memodoc (158 dPDF (29 KB) OF (25 KB)

brucejohn here is mampes take on the rns vs alpha spec analysis discrepancy

Forwarded by MelissaG TaylorR1USEPAUS on 12212005 0930 AM

Conroy Ed ltEdConroyregrn-ea ecomcomgt To

MelissaG TaylorR1USEPAUSregEPA 12152005 0416 cc PM

Subject FW Uranium Analysis Issues

Melissa

Heres Andys take on the alpha spec vs ICPMS debate along with a couple of graphs showing correlation between the two methods and MampE split data vs PRP data

Please call me on the cell phone tomorrow to discuss

Ed

Edward A Conroy PE

Project Manager

METCALF amp EDDYJAECOM

7812246880 ph

6172832942 cell

1

ed conroym-e aecom com Attachment 1

-----Original Message----shyFrom Schkuta Andrew Sent Thursday December 15 2005 358 PM To Conroy Ed Subject Uranium Analysis Issues

Ed

Please find attached a memorandum that summarizes my review of why ICP-MS and alpha spec data may not necessarily agree

ltltRad-versus-ICPMS memodocgtgt

In addition I have attached 2 preliminary charts that have the following draft information

ltltRADvsiCPMS soilsedPDFgtgt

The RADvsiCPMS soilsed chart is a draft plot of MampEs alpha spec versus ICP-MS results when data was available for both samples It shows the same general trend as the PRP data with fairly good agreement between the two analyses There appears to be a slight positive bias for alpha spec data at low concentrations but this is expected as discussed in my memo

ltltMEvsPRP ICPMSPDFgtgt

The MEvsPRP ICPMS is a draft plot of MampE vs the PRPs total uranium results as measured by ICP-MS -With the exception of a few outliers which will be further investigated there appears to be good agreement between MampEs and the PRPs data As a note the median percent difference between MampEs and the PRPs data was 1

Andrew Schkuta CHMM

Technical Specialist - Chemistry

METCALF amp EDDYJAECOM

701 Edgewater Drive

Wakefield MA 01880

(781) 224-6353

(781) 245-6293 fax

2

(See attached file Rad-versus-ICPMS Attachment 1memodoc) (See attached file RADvsiCPMS soilsedPDF) (See attached file MEvsPRP ICPMSPDF)

3

Attachment 1

Metcalf amp Eddy Inc

METCALFampEDDY

Memorandum

Date December 15 2005

To Ed Conroy

From Andrew Schkuta

Subject Uranium Analysis comparison of alpha spectrometry and ICP-MS results

File WA151 Nuclear Metals Distribution

Summary This memorandum discusses the differences between uranium analysis conducted by two different methods alpha spectroscopy (alpha spec) and inductively coupled plasma shymass spectrometry (ICP-MS) In summary the alpha spec data is expected to be biased high compared to the ICP-MS data but this is not indicative that either method is being performed incorrectly

In the September 25 2005 technical memorandum Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts (MACTEC 2005) data was presented comparing alpha spec and ICP-MS analysis for uranium analysis and it was proposed that future uranium analysis be performed by ICP-MS only For samples analyzed that were collected through Spring 2005 all samples were analyzed by ICP-MS with a subset (20) analyzed by alpha spec It was noted that there appeared to be a positive bias in the alpha spec data compared to the ICP-MS data and some concern has been voiced if analysis by ICP-MS only would meet project objectives This memorandum provides background on why the two analytical methods may yield differing results and discusses how they may impact project quality objectives

There are two primary reasons why alpha spec is expected to yield results that may be positively biased

1 Sample Preparation The sample preparation method for alpha spec is more rigorous The preparation method in alpha spec breaks down the silica mineral structure of the sample By design the ICP-MS sample preparation does not break down the mineral structure and is intended to remove all metals that could become environmentally available

2 Sample Quanitation The results for alpha spec are adjusted for analyte loss by correcting the result based on the amount of tracer recovered (yield) ICP-MS data is absolute ie data is not corrected for loss during sample preparation

2 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

These two reasons are discussed in further detail as follows

Sample Preparation

Radionuclide analysis traditionally uses a sample preparation method that includes hydrofluoric acid (HF) to break down the sample so that all of the uranium present in the sample is put into solution On the other hand standard EPA methods for metals analysis including ICP-MS do not measure the total metals present in a sample For example method SW846-3050 (EPA 1996) (as specified by the PRP in their approved QAPP for metals analysis by ICP-MS) states

This method is not a total digestion technique for most samples It is a very strong acid digestion that will dissolve almost all elements that could become environmentally available By design elements bound in silicate structures are not normally dissolved by this procedure as they are not usually mobile in the environment If absolute total digestion is required use Method 3052

As noted by the method a SW846 method is available to do a total digestion (equivalent to the alpha spec method) as method 3052 However method 3052 (EPA 1996) states (emphasis added)

Note This technique is not appropriate for regulatory applications that require the use of leachate preparations (ie Method 3050 Method 3051 Method 1311 Method 1312 Method 1310 Method 1320 Method 1330 Method 3031 Method 3040) This method is appropriate for those applications requiring a total decomposition for research purposes (ie~ geological studiesr mass balancesr analysis of Standard Reference Materials) or in response to a regulation that requires total sample decomposition

Even though the ICP-MS preparation method does not yield total results it still may conservatively measure the amount of metals that are environmentally available In the Research Project Summary Ambient Levels of Metals in New Jersey Soils prepared by the New Jersey Department of Environmental Protection (NJDEP 2003) where samples were prepared using SW846-3050 procedures it states

Metal concentrations reported in this summary were acid-extractable metals not necessarily total metal concentrations The extraction method (USEPA Method 3050) is a vigorous extraction method designed to remove all metals that could become environmentally available In practice the extraction method likely overestimates metals that could become available since the sample is refluxed with both concentrated nitric acid and hydrogen peroxide However by design the method will not extract chemicals incorporated in silica minerals as they are usually not mobile in the environment Thus concentrations reported in these studies may be lower than those indicated from analyses using methods designed to measure total metal concentrations such as x-ray fluorescence methods

Although references to uranium were not located it is expected that uranium will have some of the same type of mobility characteristics as other metals in the soil matrix The paper Ground Water Issue Behavior of Metals in Soil published by EPAs Superfund Technology Support Center for Ground Water (EPA 1992) describes how metals are distributed within soil The paper describes how metals are partitioned in various pools and how a rigorous extraction using HF is required to release metals present in the structure of primary or secondary minerals It is

3 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

expected that native metals (ie background concentrations) can be present in these structures The paper indicates that for four metals examined (cadmium copper zinc and nickel) the amount of metals that was bound in the mineral structure ranges from approximately 15-60 It would be expected that naturally occurring uranium would have similar characteristics

The National Institutes of Standards amp Technology (NIST) recognizes the differences between total digestion and method SW846-3050 and has issued Addendum to Certificates for the Certificate of Analysis (NIST 2003) for some Standard Reference Materials (SRMs) The Addendum to Certificates for SRM 2711 a Montana agricultural soil indicates that trace metals were measured at approximately 40 to 90 of their certified value when analyzed using SW846-3050 as the preparation method Even though uranium is not listed as a target analyte it is reasonable to expect that is would behave in a similar fashion

Sample Quanitation

The preparation method for alpha spec analysis includes addition of a tracer (U-232) to sample and is carried through the entire analytical scheme The amount of tracer is measured during analysis and the results for target compounds (the other uranium isotopes) are corrected for the amount of tracer recovery (yield) For example if amount of tracer measured is 50 of what was initially added to the sample the sample results are all corrected by dividing by 050 In terms of method quality assurancequality control (QAQC) the PRP QAPP indicates that tracer recovery in the range of 15-125 is considered acceptable and the analytical system is considered in control In other words up to 85 of the compound can be lost during sample preparation and analysis and the data is considered meeting all QAQC objectives This method of analysis is the equivalent to the technique of isotope dilution which is used for some non-radiological environmental analyses (eg EPA Method 1625 Semivolatile Organic Compounds by Isotope Dilution GCMS) Analyses using this technique although considered the most accurate are not widely employed due to the complexity and expense of using the isotope dilution technique

ICP-MS analysis measures the concentration of analyte with no correction for potential losses However QAQC procedures are in place to make sure the losses are not considered significant In addition to laboratory control samples matrix spike analysis measures the recovery for the target analyte that was over-spiked into the sample The PRP QAPP states that matrix spike recovery should be within the range of 75-125 Therefore up to 25 of the analyte could be lost and the method has still met the QAQC objectives If recovery is less than this then validation procedures acknowledge that the data may be biased low and sample results are flagged as estimated

Therefore because the alpha spec data is corrected for analyte loss during the analytical process while ICP-MS data is not some amount of positive bias is expected to be present in alpha spec data when compared to ICP-MS data

Considerations

In evaluating which analysis is appropriate for samples from the project neither method is inherently better than the other The following factors should be considered

1 What is being measured ICP-MS measures mass directly and data can be mathematically converted to activity Alpha spec measures activity directly and can be mathematically converted to mass

4 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

2 What are the objectives of the data in terms of measuring total versus environmentally available The ICP-MS method specified in the PRP QAPP measures uranium concentrations in the same manner that EPA has measured other metals concentrations Should uranium mass be measured in a typical way that will be used for measuring other metals of concern at the site (eg copper beryllium arsenic etc)

3 If ICP-MS is more cost effective that alpha spec should the option of using preparation method SW846-3052 be considered This should yield total uranium data similar to alpha spec analysis while still using the efficiencies of ICP-MS analysis

4 What units will the cleanup level for the site be in If its in mass units then ICPshyMS analysis may be more appropriate If its in activity units then alpha spec analysis may be warranted

References

Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts Technical Memorandum from Heath Downey to Bruce Thompson de maxim is inc September 26 2005 (MACTEC 2005)

Test Methods for Evaluating Solid Waste PhysicalChemical Methods SW-846 Third Edition with updates EPA 1996

Ambient Levels of Metals in New Jersey Soils Environmental Assessment and Risk Analysis Element Research Project Summary Division of Science Research and Technology State of New Jersey Department of Environmental Protection May2003

Ground Water Issue Behavior of Metals in Soil EPA Superfund Technology Support Center for Ground Water EPA540S-92018 October 1993

Certificate of Analysis Standard Reference Material 2711 Montana Soil National Institute of Standards amp Technology July 18 2003

bull bull

bull bull bullbull bull

10000

MampE SoilSediment Results Total Uranium mgKg

bull

bull - ~-~-----~~---~~-~--~~1000

en aE

I I bull Total a u

I -

bull bull 100 1shy -middot- -middot bullbull bull

Ybull bull

10 100 1000 10000

Alpha Spec

Attachm

ent 1

10 -

100000

TOTAL URANIUM BY ICP-MS mgKg

bull

bull 10000

I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 4: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

de maximis inc

The ICPMS digestion procedure is designed to evaluate metals that could become ldquoenvironmentallyrdquo available whereas the alpha spectroscopy digestion is an extremely aggressive technique where complete mineral dissolution is performed As a result the generally higher uranium concentrations reported for the alpha spectroscopy analyses are not unexpected and may be explained by the liberation of geologically-bound uranium during digestion

The following information describes the methodologies and differences between the two analytical methods

bull Preparation homogenization and digestion ndash laboratory SOPs for the RIFS (QAPP 2004) are presented in Attachment 2 for these techniques The alpha spectroscopy preparation technique (SOP GL-RAD-015) was adapted from the DOE HASL Manual and uses a larger aliquot of sample and dries grinds ashes the sample prior to acid digestion The alpha spectroscopy analysis uses complete dissolution (HNO3 and HF) The ICPMS technique (SOP GL-MA-Eshy009) follows the EPA digestion method approved for the RIFS (QAPP 2004) and includes homogenization of the wet sample (smaller aliquot) and digestion using hot acid (HNO3 and H2O2)

bull Sample Quantitation - alpha spectroscopy method quantitation procedure includes adjustment of instrument results based on tracer R (percent yield is factored into reported result) ICPMS method quantitation procedure does not include applying correction factors for any bias observed during analysis (ie an absolute value is reported)

bull The ICPMS hot acid digestion technique used for the RIFS is a standard EPA digestion technique used at hundreds of CERCLA sites and is a widely accepted and aggressive method that is appropriate for a wide range of metals including uranium and thorium Furthermore as documented in the laboratory ICPMS SOP (GL-MA-E-014) uranium isotopes and thorium are suitable metals for analysis

bull As shown in Attachment 1 there is a very strong comparison between uranium split-sample results reported by de maximis inc and MampE which demonstrates the representativeness and usability of ICPMS data for the Site

Another factor to consider as part of the comparison of analytical methods is the resolution of uranium by each method The primary isotope of natural and depleted uranium is U-238 which has a low specific activity (ie it takes a large amount of mass to increase the amount of radioactivity) For example each increase in radioactivity by 1 pCig of U-238 results in an increase in mass of 3 mgkg Due to the fact that ICPMS has better resolution and sensitivity at lower mass concentrations it was chosen (and approved by EPA) as the quantitative analytical method for uranium for the RIFS

4

de maximis inc

EPA metals digestion procedures for all analyses (whether ICP ICPMS cold vapor) do not ldquounder-leachrdquo samples rather digestion procedures are designed to evaluate ldquoenvironmentally availablerdquo concentrations As such uranium and thorium would be analyzed as for other potential metal contaminants at the site (Cu Be Pb Hg etc)

While it is in fact possible that some disposed uranium could resist the hot nitric digestion it would also certainly resist leaching by rainwater infiltration or groundwater flushing and be less biologically available if ingested

To assess laboratory performance of the QAPP-approved methods for the analysis of uranium and thorium for the project performance evaluation (PE) samples have been submitted to the laboratory As part of the Phase IA and Phase 1B investigation programs blind PE samples (quality control standard that is of a composition and concentration not known to the laboratory) were submitted to the laboratory for analysis Results of the analyses and comparison to vendor spike concentrations are provided below

Investigation Analysis Parameter Laboratory Result Certified Value (Vendor Spike)

QC PALs

Phase 1A ICPMS Uranium (total)

32 26 24 mgkg1 22 mgkg NA

Phase 1A ICPMS Thorium 229 159 162 110 mgkg1

131 mgkg NA

Phase 1A Alpha Spec

Uraniumshy238

313 pCig 339 pCig NA

Phase 1A Alpha Spec

Uraniumshy234

34 pCig 339 pCig NA

Phase 1A Alpha Spec

Uraniumshy235

0252 U pCig2 016 pCig NA

Phase 1A Alpha Spec

Thorium-230 261 pCig 134 pCig NA

Phase 1B Alpha Spec

Uranium (total)

1184 pCiL 117 pCiL 852 ndash 138 pCiL

Phase 1B ICPMS Uranium 95 mgkg 126 mgkg 834 ndash 144 mgkg

Phase 1B ICPMS Uranium 606 ugL 635 ugL 525 ndash 745 ugL

1 = results provided for multiple runs performed by the laboratory 2 = not detected laboratory minimum detectable activity is greater than amount present in PE sample NA = not available

As seen in this table the laboratory provided acceptable results (ie results generally showed good agreement to vendor certified values and results were reported within QC Performance Acceptance Limits of the vendor result (when applicable) as determined by peer laboratory measurements) indicating the QAPP methods used for uranium and thorium analysis are capable of providing usable results for the RIFS Vendor certified value is a ldquomade tordquo value as determined by volumetric andor gravimetric measurements of National Institute of Standards and Technology (NIST) Standard

5

de maximis inc

Reference Materials or NIST traceable reference material andor analytically verified by the vendor Vendor Performance Acceptance Limits are based on actual historical data collected in the vendors proficiency testing program The limits reflect any inherent biases in the methods used to establish the limits and closely approximate a 95 confidence interval of the performance that experienced laboratories should achieve using accepted environmental methods

An additional statistical evaluation of the analysis of uranium by ICPMS and alpha spectroscopy was performed using linear regression Linear regression mathematically approximates the relationship between two variables and provides an equation to predict a dependent variable from an independent variable Since ICPMS was selected as the primary analytical method for uranium is was assigned as the independent variable (x) and alpha spectroscopy was assigned as the dependent variable (y)

Furthermore the relationship between these two methods was anticipated to be linear such that y = x ideally so the y-intercept was set to equal zero for linear regression During the Fall 2004 RI phase of the investigation 172 soil and sediment samples were collected and analyzed by both analytical methods Of these samples 3 were determined to be from the super-depleted subset by ICPMS and were excluded from linear regression as the results of these samples are questionable

The results of linear regression for this data set are an estimated model of y (alpha spectroscopy) = 10604 x (ICPMS) with an R2 value of 09049 R2 measures the strength of the linear regression model with a value of 1 indicating a perfect linear fit so the results indicate a very strong fit In addition the 95 confidence interval for the slope ranges from 10080 to 11128 which indicates that linear regression modeled alpha spectroscopy results are generally 6 higher than ICPMS results with a range of 1 to 11 (95 confidence) The results of the linear regression analysis are shown in the attached Figure 2 (note that the scale is logarithmic in order to allow better display of all the data points across the range of results) From Figure 2 it is evident that there is substantial scatter in the low concentration range (less than 10 mgkg) and if these data points are removed from the linear regression evaluation there is no significant change in the results since the y-intercept was set to equal zero

These linear regression modeling results provide strong evidence that when the entire population of uranium analysis results for both ICPMS and alpha spectroscopy are considered there is good agreement between the two methods The relative difference is not significant when compared to the expected accuracy of the analytical techniques (+25)

In response to the comment regarding use of alpha spectroscopy values rather than ICPMS values to make remedial decisions it is clear that the commenter is not sufficiently familiar with the CERCLA investigation and remediation process The purpose of the remedial investigation (RI) is to collect data to evaluate the potential risks to human health and the environment associated with chemicals of potential concerns (COPCs) at the Site The data from the two methods show good overall correlation but as indicated there is a discrepancy if only the 1-50 ppm U-238 results are considered

6

de maximis inc

It is important to understand how the RI data will be evaluated and interpreted moving forward in the CERCLA process with respect to the concern that some areas might be missed for remediation The RI data set will be used to derive human exposure point concentrations (EPCs) for COPCs within the various media and exposure areas at the Site EPCs will be derived in accordance with EPA guidance which generally use 95 upper confidence level concentrations for soil analytes For uranium this will be performed using the ICPMS data set As a component of the risk assessment uncertainties associated with the analytical data are evaluated evaluation of alpha spectroscopy outliers as well as evaluation of the effect of uranium enrichment (eg ranging from 02 to 07) on risk estimates will be part of this process

If based on this process a given exposure areamedia has risks that EPA deems to be unacceptable (generally gt 10-4 cancer risk or non-cancer hazard index gt1 based on the cumulative risk from all COPCs in a given mediumexposure area) then preliminary remediation goals (PRGs) will be developed for the COPCs that contribute most substantially to the total risk In turn the PRGs will then be used to evaluate areas and volumes for which remedial alternatives will be considered in the Feasibility Study (FS) Areas for remediation will consider a variety of factors when determining the overall dimensions and they typically will be drawn to include RI data points that are equal to or greater than the PRGs

EPA will then select final remediation goals in the Record of Decision (ROD) The Remedial DesignRemedial Action (RDRA) process occurs after the ROD During the RD areas and volumes for remediation will be revisited based on the final remediation goals typically resulting in identification of a minimum target areavolume that will be remediated As part of the RD process a Demonstration of Compliance Plan will be developed This document sets forth how the contractor will demonstrate that all targeted media has been successfully remediated For a site like NMI we would expect that different instrumentation analytical methods may be used on Site to evaluate the effectiveness of remediation during the RA (eg gamma spectroscopy or XRF which provides faster results than ICPMS or alpha spectroscopy) followed by confirmatory sampling sent to an off-site laboratory Correlation of these instrument results to laboratory analytical methods will be presented at the appropriate time Given this process (CERCLA) and the multiple reviews by EPA State of Massachusetts agencies and public groups it is very unlikely that any areas with unacceptable risk levels will be missed for remediation

With respect to uranium background concentrations the total uranium values calculated using alpha spectroscopy values is correct in the CREW Technical Note (2006) however this background value is based on a theoretical calculation While this calculation is correct de maximis inc (and EPA via MampE split samples) have shown total uranium (reflected in the background evaluation) directly measured by ICPMS is lower due to inherent differences between the two methods and the fact that ICPMS is measuring ldquoenvironmentally availablerdquo concentrations Furthermore the background concentration at the Site is not likely to play a significant role since risk-based remediation levels greater than background are expected to be established for Site clean-up

7

de maximis inc

Lastly the intent for referencing the urine study was to show an independent evaluation that also concludes ICPMS has lower sensitivity for uranium and a greater ability to distinguish between natural depleted and enriched uranium Further it supports the argument that some variability between ICPMS and alpha spectroscopy analysis is not unexpected

In conclusion the ICPMS analysis provides lower results as compared to alpha spectroscopy analysis results however as discussed in this response (and described in detail in MampE uranium evaluation memo) neither method is in error

bull Background uranium concentrations determined from actual ICPMS measurements show environmentally available U background is approximately 13 ppm (EPA MampE-split samples confirm this)

bull lsquoSuper-depletedrsquo points are anomalies and these samples will be recollected and reanalyzed however remaining total U concentrations are usable and ICPMS results generated for the RIFS are reliable

bull Urine analysis referenced in the Technical Memorandum is applicable in that it supports the use of ICPMS for evaluating presence of U235 and calculating enrichments (for future samples once alpha spectroscopy analysis is discontinued)

bull Alpha spectroscopy analysis analyses for the Site were intended to verify enrichment to ensure that no enriched uranium was missed by ICPMS and to confirm signature of uranium and thorium at the site This has been done therefore future alpha spectroscopy analysis is not needed

Response to Section 3 ndash Using enrichment rather than total U for determination of contamination

Using total uranium by mass rather than enrichment to define nature and extent of contamination is being performed at this Site because the determination of enrichment at total uranium concentrations in the background range (eg 075 ndash 25 mgkg) is uncertain particularly for alpha spectroscopy analysis methods where the total propagated uncertainty of measurements is nearly equal to the actual instrument reading This is clearly demonstrated in the scatter observed in data plots provided by CREW (2006) for total uranium in this range de maximis inc concurs that at 5 mgkg a large percentage of the mass should be depleted uranium and for the purposes of defining the nature and extent of contamination based solely on isotopic composition a value of 5 would be inappropriate However the RI is using a value of 21 mgkg as a threshold for the definition of the nature and extent of contamination The nature and extent of uranium like all substances at the Site is being defined by risk-based levels (background levels are only used when background values are higher than risk-based levels) Therefore for total uranium concentrations less than the RI screening level the

8

de maximis inc

source of the uranium (whether depleted or natural uranium) is not relevant For risk assessment purposes all uranium will be considered and the contribution to risk associated with background constituents of concern will be considered after the total risk is calculated

Measurements of environmental data rarely have a normal (Gaussian) distribution and typically are lognormal distributions (Gilbert 1987) EPA guidance indicates that ldquoit is unusual to encounter environmental data sets that are normally distributedrdquo (EPA 2002) and ldquoenvironmental data commonly exhibit frequency distributions that are non-negative and skewed with heavy or long right tailsrdquo (EPA 2000) The composition of uranium (natural or depleted) will be used during risk assessment but it has limited usefulness otherwise However the isotopic composition of uranium will be one of the issues considered during data evaluation to avoid the sort of issues pointed out in the CREW Technical Note

Response to Section 4 - Thorium evaluation

de maximis inc re-iterates that the Technical Memorandum shows thorium-232 and thorium-228 concentrations are roughly equal (and considered within the analytical variability for alpha spectroscopy) and that thorium-230 at the Site is not related to Site activities Furthermore uranium-232 (from recycled uranium) is not present at the site as it is primarily a concern with enriched uranium Even if present it would not be in significant enough quantities to impact thorium-228 concentrations

Refer to previous discussion pertaining to ICPMS versus alpha spectroscopy analysis information Also note that ICPMS instrument sensitivity for thorium is very high due to the high mass to charge ratio (ie it is a good responder) Naturally occurring thorium is 100 Th-232 No data quality issues were observed during data validation and data has been deemed usable

Lastly thorium was used at the Site but at significantly lower amounts when compared to uranium Therefore low concentrations (within background concentration range) of thorium are not unexpected and historical data have demonstrated this

Response to Section 5 - Other radionuclides

There is no way to know the exact pedigree of the depleted uranium processed at the Site or to determine which portion of the depleted uranium might have come from reprocessed fuel rods and therefore contain small amounts of reactor generated radionuclides Studies of depleted uranium have shown that byproduct radionuclides in uranium are present at very low levels In addition dose assessments of these trace contaminants have shown that the radiological dose from these radionuclides is insignificant compared to that from uranium These conclusions were documented in a January 2000 Department of the Army analysis of transuranics and other contaminants in depleted uranium armor (Attachment 3) Of particular relevance is the fact that the Department of the Army analysis took samples of DU billets from Starmet for this study The Army concluded that ldquoThe levels of transuranics elements (TRUs) and fission

9

de maximis inc

products (FP) are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is saferdquo If the TRUs and FP contribute less than a one percent increase in dose for DU metal then they would contribute an even more negligible dose at the 10s to 1000s of mgkg of DU present in NMI Site media With respect to the HTDR analysis during the RI historical data indicated that the holding basin and former drain outfall regions of the bog were the most likely areas to have significant concentrations of uranium therefore the RI work plan identified locations to be sampled in these areas (FSP 2004) Further the approach outlined in the RI was to evaluate for HTDR in samples with elevated concentrations of uranium not necessarily the highest concentrations There were other AOIs where uranium was detected in elevated concentrations but no sample was collected for HTDR However since the holding basin was in use for the vast majority of time that uranium was actively processed at the site and no HTDR were detected in the six samples from this area it is unlikely that HTDR would be identified in other areas of the site

Samples for HTDRs were collected according to the RI (FSP 2004) such that 3 of the samples from AOI 1 were above the water table and 3 of the samples were from below the water table one sample was collected from each of the AOI 6 former drain outfall areas of the bog and one sample collected from the buried drum excavation (AOI 2) These HTDR samples are representative for the areas investigated (AOI 1 AOI 2 and AOI 6) and of the site based on the distribution of total uranium results described below

bull Total uranium in the HTDR sample locations from AOI 1 ranged from 38 to 511 mgkg while total uranium for all samples in AOI 1 ranged from 11 to 672 mgkg Only 25 samples from all samples in AOI 1 had total uranium concentrations greater than 100 mgkg and 4 of the 6 HTDR samples were from this subset

bull Total uranium in the HTDR sample locations from AOI 6 were 25 mgkg at former drain outfall 1 and 4550 mgkg at drain outfall 2 while total uranium for all samples in AOI 6 ranged from 16 to 4550 mgkg Only 10 samples from all samples in AOI 6 had total uranium concentrations greater than 100 mgkg and one HTDR sample was from the maximum total uranium sample location

bull The HTDR sample from AOI 2 drum excavation bottom had a total uranium concentration of 18 mgkg while total uranium for all excavation sides and bottom ranged from 14 to 49 mgkg

Only a few samples have uranium concentrations greater than 1000 mgkg and 6 of the 8 are from samples deemed as lsquosuper-depletedrsquo These 6 locations will be re-sampled as part of Phase 1C since the ICPMS analysis results are considered an anomaly Sporadic low detections of HTDRs do not necessarily indicate the presence of HTDRs as false positive results are common for these analyses Although some HTDRs have been detected in the past they are sporadic (not all samples have detections) and do not indicate the presence of HTDRs at significant levels

10

de maximis inc

It is also important to note that the historical data provided in the RI (PSOP 2005) was used to assist in design of the RI (FSP 2004) Data were compiled from many sources of information and represent samples collected analyzed and reported using various sample collection procedures laboratory analytical methods and data reporting conventions Data compilation activities involved both manual hand-entries and electronic entries when applicable Based on a review of the electronic database severe data quality limitations were noted including unknown analytical quality and validation levels missing location coordinates inconsistent media codes unknown detection limits and radionuclide uncertainty measurements inconsistent andor incorrect reporting of radionuclides using mass-based and activity-based units These limitations were considered significant during the scoping of the RIFS Work Plan therefore the historical data were used primarily to evaluate the spatial distribution of parameters and magnitude of detected concentrations The limitations of the historical data were used to provide the basis for identifying data gaps and scoping the RIFS As such the FSP included analysis for HTDRs at the areas of the Site known to be impacted by historical practices

With respect to comments provided on U-232 and U-236 these uranium isotopes are produced during the fission of uranium (as are plutonium americium and the rest of the HTDRs) Since none of these other fission products have been detected at the Site it is not credible that either uranium-232 or uranium-236 would be present

Response to Section 6 - Radium

The historical radium in groundwater data is suspect for several reasons including the previous discussion on historical data provided in the RI (PSOP 2005) Groundwater samples were collected from 13 wells at the Site (single sampling event) using bailers as part of this investigation All samples were screened for radioactivity by gross alpha beta analysis Only one sample exceeded the action level (gt 10 pCil gross alpha) with results of 50 pCil gross alpha and 55 pCil gross beta and it was analyzed for isotopic uranium Ra-226 and Ra-228 The results were 74 pCil U-234 025 pCil U-235 037 pCil U-238 12 pCil Ra-226 and 7 pCil Ra-228 These values are from Table 10 of the report and it appears that the Ra-226 value was incorrectly transcribed in the CREW Technical Note as 37 pCil The isotopic uranium results do not match those expected as U-234 is elevated and results for U-235 and U-238 are almost equal Therefore the analytical results for this sample are questionable Furthermore all the rest of the ground water samples had gross alpha results of 3 pCil or less indicating that there are no significant amounts of radium in groundwater

In addition a limited literature review was performed to assess the potential for leaching of radium from subsurface solids by nitrate at the NMI Site The literature demonstrates that radium may be found in a handful of aqueous phase complexes The predominant species of radium in aqueous solution is the divalent cation Ra2+ but other complexes such as RaOH+ RaCl+ RaCO3 and RaSO4 may also be found (Beneš et al 1982 Beneš 1984) Radium appears to be the least likely of all the alkali earth metals (beryllium magnesium calcium strontium barium and radium) to form aqueous phase

11

de maximis inc

complexes (Kirby and Salutsky 1964) As such we are unclear as to the origin of Dr Landreacutersquos statement that ldquonitrates are known to influence greatly radium leaching from the soil into the groundwaterrdquo (Landreacute 2006) The salt Ra(NO3)2 is highly soluble (245 g100 g H2O Kirby and Salutsky 1964) but we have not found evidence to suggest that the species Ra(NO3)+ or Ra(NO3)2 exist in aqueous solution Lauria et al (2004) present radionuclide and water chemistry data from groundwater samples taken from the Buena Lagoon in Brazil A plot of Ra-226 and Ra-228 concentrations versus nitrate concentration is presented in Figure 1 These data demonstrate that both Ra-226 and Ra-228 are almost completely uncorrelated with nitrate at this site as evidenced by correlation coefficients (r2) of approximately 005 and 004 respectively Szabo et al (1997) also report that no correlation exists between nitrate and Ra-226 in 17 groundwater samples taken from the Kirkwood-Cohansey aquifer system in New Jersey although Ra-228 appears to be weakly correlated with nitrate in the same data set (r2 = 018)

Figure 1 Correlation between aqueous radium and nitrate (Lauria et al 2004)

r 2 = 005

r 2 = 004

1E-10 1E-09 1E-08 1E-07 1E-06 1E-05 1E-04

0 5 10 15 20 25

Nitrate (mgL)

Ra

(microg

L)

Ra-226 Ra-228 Linear (Ra-226) Linear (Ra-228)

The mobility of radium in the subsurface appears to be controlled primarily by sorption and precipitation (Lauria et al 2004) High values of aqueous phase radium have been observed in groundwater samples with either increased salinity or reduced pH (Kraemer and Reed 1984 Lauria and Godoy 2002) Demir (1989) demonstrated the increased mobility of radium in acidic environments due to the reduced sorptive capacity of sediments at low pH Figure 2 is a plot of radium concentration versus sample pH from the same data set (Lauria et al 2004)

12

de maximis inc

Figure 2 Correlation between aqueous radium and pH (Lauria et al 2004) 226Ra = -4E-06pH + 3E-05

r2 = 058

228Ra = -5E-08pH + 4E-07 r2 = 032

1E-10 1E-09 1E-08 1E-07 1E-06 1E-05 1E-04

3 4 5 6 7

pH

Ra

(microg

L)

Ra-226 Ra-228 Linear (Ra-226) Linear (Ra-228)

Here a stronger correlation between Ra-226 concentration and pH exists (r2 = 058) along with an observable trend of decreasing aqueous phase radium with increasing pH Szabo et al (1997) also report a negative correlation (r2 = 030) between Ra-228 and pH but no significant correlation between Ra-226 and pH Divalent cations such as calcium and magnesium compete with radium for a limited number of surface complexation and adsorption sites (Szabo et al 1997) As lime (a mixture potentially containing calcium oxide calcium hydroxide and calcium carbonate) was co-disposed with uranium and nitric acid at the Holding Basin nitrate may serve as a site-specific tracer for conditions that may lead to the release of adsorbed or surface-complexed radium from the solid phase although no comparable calcium ldquoplumerdquo has been defined unlike nitrate in groundwater Therefore we suggest that Dr Landreacutersquos statement could be qualified by stating that low pH values are known to influence the leaching of radium from aquifer solids However as discussed above we do not see a compelling reason to add radium to the analytical list at this time

Please contact me if you have any questions

Sincerely

Bruce Thompson

cc Respondents Bennie Underwood PE de maximis inc inc Nelson Walter PE MACTEC Inc Peter Zeeb PhD PG GeoSyntec Consultants Inc

13

de maximis inc

Figure 1 ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

Figure 2 Comparison of Alpha Spectroscopy and ICP-MS Uranium Analysis

Attachment 1 Metcalf amp Eddy Memorandum ndash Uranium Analysis comparison of alpha spectrometry and ICP-MS results

Attachment 2 QAPP SOPs for Alpha Spectroscopy and ICPMS Sample Preparation

Attachment 3 Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor Department of the Army January 19 2000

References

Beneš P Migration of radium in the terrestrial hydrosphere The behavior of radium in waterways and aquifers Technical Reports Series No 310 International Atomic Energy Agency IAEA-TECDOC-301 Vienna Austrai pp 119-173 1984

Beneš P M Obdržaacutelek M Čejchanovaacute The physicochemical forms of tracers of radium in aqueous solutions containing chlorides sulfates and carbonates Radiochem Radioanal Letters 50(4) pp 227-242 1982

CREW 2006 ldquoComments on Radiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Technical Note No 2006-01 January 6 2006

Demir I Temporal variations and sources of Pb Co Cu Ni Fe and Mo in shallow ground water of the McDonalds Branch basin Lebanon State Forest Burlington County New Jersey New Jersey Geological Survey Report 20 19 p 1989

EPA 2000 Guidance for Data Quality Assessment Practical Methods for Data Analysis EPA QAG-9 July 2000

EPA 2002 Guidance for Comparing Background and Chemical Concentrations in Soil for CERCLA Site EPA 540-R-01-003 September 2002

FSP 2004 Field Sampling Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Gilbert 1989 Statistical Methods for Environmental Pollution Monitoring Richard O Gilbert 1987

Kirby H W M L Salutsky The Radiochemistry of Radium National Academy of Sciences ndash National Research Council NAS-NS 3057 205 p 1964

14

de maximis inc

Kramer T F and D F Reid The occurrence and behavior of radium in saline formation water of the US Gulf Coast Region Isotope Geoscience 2 pp 153shy174 1984

Landreacute V Comments on ldquoRadiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Nuclear Metals Superfund Site CREW Technical Note 2006-01 January 6 2006

Lauria D C and J M O Godoy Abnormal high natural radium concentration in surface waters Journal of Environmental Radioactivity 61(2) pp 27-37 2002

Lauria D C R M R Almeida O Sracek Behavior of radium thorium and uranium in groundwater near the Buena Lagoon in the Coastal Zone of the State of Rio de Janeiro Brazil Environmental Geology 47 pp 11-19 200

MACTEC 2005 Radiological Review of Fall 2004 RI Analytical Data Nuclear Metals Superfund Site Concord Massachusetts September 26 2004

MampE 2005 Memorandum from Andrew Schkuta to Ed Conroy Subject Uranium Analysis comparison of alpha spectrometry and ICPMS results December 15 2005

PSOP Project Summary and Operations Plan Nuclear Metals Superfund Site Concord Massachusetts April 15 2005

QAPP 2004 Quality Assurance Project Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Szabo Z D E Rice C L MacLeod and T H Barringer Relation of distribution of radium nitrate and pesticides to agricultural land use and depth Kirkwood-Cohansey aquifer system New Jersey Coastal Plain 1990-1991 US Geological Survey Water-Resources Investigations Report 96-4165A 119 p 1997

15

bull

bull bull

bull bull

bull bull bull

bull bull

bull bull

bull

500

1500

Figure 1

Con

cent

ratio

n (m

gkg

)

ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

1000

0

0 5 10 15 20 25 30 35 40

Samples

Figure 2

Alp

ha S

pect

osco

py (m

gkg

)

Fall 2004 Total Uranium No Super-Depleted

1000

100

10

1

y = 10604x R2 = 09049

1 10 100 1000

ICPMS (mgkg)

Fall 2004 Super-Depleted Linear Regression

Attachment 1Connolly Jayme

From Taylor Mel issagepam ail epa gov Sent Wednesday December 21 2005 932AM To brucetdemaximiscom jhuntdemaximiscom Subject Fw Uranium Analysis Issues

Rad-versus-ICPMS RADvsiCPMS_soilse MEvsPRP_ICPMSP memodoc (158 dPDF (29 KB) OF (25 KB)

brucejohn here is mampes take on the rns vs alpha spec analysis discrepancy

Forwarded by MelissaG TaylorR1USEPAUS on 12212005 0930 AM

Conroy Ed ltEdConroyregrn-ea ecomcomgt To

MelissaG TaylorR1USEPAUSregEPA 12152005 0416 cc PM

Subject FW Uranium Analysis Issues

Melissa

Heres Andys take on the alpha spec vs ICPMS debate along with a couple of graphs showing correlation between the two methods and MampE split data vs PRP data

Please call me on the cell phone tomorrow to discuss

Ed

Edward A Conroy PE

Project Manager

METCALF amp EDDYJAECOM

7812246880 ph

6172832942 cell

1

ed conroym-e aecom com Attachment 1

-----Original Message----shyFrom Schkuta Andrew Sent Thursday December 15 2005 358 PM To Conroy Ed Subject Uranium Analysis Issues

Ed

Please find attached a memorandum that summarizes my review of why ICP-MS and alpha spec data may not necessarily agree

ltltRad-versus-ICPMS memodocgtgt

In addition I have attached 2 preliminary charts that have the following draft information

ltltRADvsiCPMS soilsedPDFgtgt

The RADvsiCPMS soilsed chart is a draft plot of MampEs alpha spec versus ICP-MS results when data was available for both samples It shows the same general trend as the PRP data with fairly good agreement between the two analyses There appears to be a slight positive bias for alpha spec data at low concentrations but this is expected as discussed in my memo

ltltMEvsPRP ICPMSPDFgtgt

The MEvsPRP ICPMS is a draft plot of MampE vs the PRPs total uranium results as measured by ICP-MS -With the exception of a few outliers which will be further investigated there appears to be good agreement between MampEs and the PRPs data As a note the median percent difference between MampEs and the PRPs data was 1

Andrew Schkuta CHMM

Technical Specialist - Chemistry

METCALF amp EDDYJAECOM

701 Edgewater Drive

Wakefield MA 01880

(781) 224-6353

(781) 245-6293 fax

2

(See attached file Rad-versus-ICPMS Attachment 1memodoc) (See attached file RADvsiCPMS soilsedPDF) (See attached file MEvsPRP ICPMSPDF)

3

Attachment 1

Metcalf amp Eddy Inc

METCALFampEDDY

Memorandum

Date December 15 2005

To Ed Conroy

From Andrew Schkuta

Subject Uranium Analysis comparison of alpha spectrometry and ICP-MS results

File WA151 Nuclear Metals Distribution

Summary This memorandum discusses the differences between uranium analysis conducted by two different methods alpha spectroscopy (alpha spec) and inductively coupled plasma shymass spectrometry (ICP-MS) In summary the alpha spec data is expected to be biased high compared to the ICP-MS data but this is not indicative that either method is being performed incorrectly

In the September 25 2005 technical memorandum Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts (MACTEC 2005) data was presented comparing alpha spec and ICP-MS analysis for uranium analysis and it was proposed that future uranium analysis be performed by ICP-MS only For samples analyzed that were collected through Spring 2005 all samples were analyzed by ICP-MS with a subset (20) analyzed by alpha spec It was noted that there appeared to be a positive bias in the alpha spec data compared to the ICP-MS data and some concern has been voiced if analysis by ICP-MS only would meet project objectives This memorandum provides background on why the two analytical methods may yield differing results and discusses how they may impact project quality objectives

There are two primary reasons why alpha spec is expected to yield results that may be positively biased

1 Sample Preparation The sample preparation method for alpha spec is more rigorous The preparation method in alpha spec breaks down the silica mineral structure of the sample By design the ICP-MS sample preparation does not break down the mineral structure and is intended to remove all metals that could become environmentally available

2 Sample Quanitation The results for alpha spec are adjusted for analyte loss by correcting the result based on the amount of tracer recovered (yield) ICP-MS data is absolute ie data is not corrected for loss during sample preparation

2 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

These two reasons are discussed in further detail as follows

Sample Preparation

Radionuclide analysis traditionally uses a sample preparation method that includes hydrofluoric acid (HF) to break down the sample so that all of the uranium present in the sample is put into solution On the other hand standard EPA methods for metals analysis including ICP-MS do not measure the total metals present in a sample For example method SW846-3050 (EPA 1996) (as specified by the PRP in their approved QAPP for metals analysis by ICP-MS) states

This method is not a total digestion technique for most samples It is a very strong acid digestion that will dissolve almost all elements that could become environmentally available By design elements bound in silicate structures are not normally dissolved by this procedure as they are not usually mobile in the environment If absolute total digestion is required use Method 3052

As noted by the method a SW846 method is available to do a total digestion (equivalent to the alpha spec method) as method 3052 However method 3052 (EPA 1996) states (emphasis added)

Note This technique is not appropriate for regulatory applications that require the use of leachate preparations (ie Method 3050 Method 3051 Method 1311 Method 1312 Method 1310 Method 1320 Method 1330 Method 3031 Method 3040) This method is appropriate for those applications requiring a total decomposition for research purposes (ie~ geological studiesr mass balancesr analysis of Standard Reference Materials) or in response to a regulation that requires total sample decomposition

Even though the ICP-MS preparation method does not yield total results it still may conservatively measure the amount of metals that are environmentally available In the Research Project Summary Ambient Levels of Metals in New Jersey Soils prepared by the New Jersey Department of Environmental Protection (NJDEP 2003) where samples were prepared using SW846-3050 procedures it states

Metal concentrations reported in this summary were acid-extractable metals not necessarily total metal concentrations The extraction method (USEPA Method 3050) is a vigorous extraction method designed to remove all metals that could become environmentally available In practice the extraction method likely overestimates metals that could become available since the sample is refluxed with both concentrated nitric acid and hydrogen peroxide However by design the method will not extract chemicals incorporated in silica minerals as they are usually not mobile in the environment Thus concentrations reported in these studies may be lower than those indicated from analyses using methods designed to measure total metal concentrations such as x-ray fluorescence methods

Although references to uranium were not located it is expected that uranium will have some of the same type of mobility characteristics as other metals in the soil matrix The paper Ground Water Issue Behavior of Metals in Soil published by EPAs Superfund Technology Support Center for Ground Water (EPA 1992) describes how metals are distributed within soil The paper describes how metals are partitioned in various pools and how a rigorous extraction using HF is required to release metals present in the structure of primary or secondary minerals It is

3 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

expected that native metals (ie background concentrations) can be present in these structures The paper indicates that for four metals examined (cadmium copper zinc and nickel) the amount of metals that was bound in the mineral structure ranges from approximately 15-60 It would be expected that naturally occurring uranium would have similar characteristics

The National Institutes of Standards amp Technology (NIST) recognizes the differences between total digestion and method SW846-3050 and has issued Addendum to Certificates for the Certificate of Analysis (NIST 2003) for some Standard Reference Materials (SRMs) The Addendum to Certificates for SRM 2711 a Montana agricultural soil indicates that trace metals were measured at approximately 40 to 90 of their certified value when analyzed using SW846-3050 as the preparation method Even though uranium is not listed as a target analyte it is reasonable to expect that is would behave in a similar fashion

Sample Quanitation

The preparation method for alpha spec analysis includes addition of a tracer (U-232) to sample and is carried through the entire analytical scheme The amount of tracer is measured during analysis and the results for target compounds (the other uranium isotopes) are corrected for the amount of tracer recovery (yield) For example if amount of tracer measured is 50 of what was initially added to the sample the sample results are all corrected by dividing by 050 In terms of method quality assurancequality control (QAQC) the PRP QAPP indicates that tracer recovery in the range of 15-125 is considered acceptable and the analytical system is considered in control In other words up to 85 of the compound can be lost during sample preparation and analysis and the data is considered meeting all QAQC objectives This method of analysis is the equivalent to the technique of isotope dilution which is used for some non-radiological environmental analyses (eg EPA Method 1625 Semivolatile Organic Compounds by Isotope Dilution GCMS) Analyses using this technique although considered the most accurate are not widely employed due to the complexity and expense of using the isotope dilution technique

ICP-MS analysis measures the concentration of analyte with no correction for potential losses However QAQC procedures are in place to make sure the losses are not considered significant In addition to laboratory control samples matrix spike analysis measures the recovery for the target analyte that was over-spiked into the sample The PRP QAPP states that matrix spike recovery should be within the range of 75-125 Therefore up to 25 of the analyte could be lost and the method has still met the QAQC objectives If recovery is less than this then validation procedures acknowledge that the data may be biased low and sample results are flagged as estimated

Therefore because the alpha spec data is corrected for analyte loss during the analytical process while ICP-MS data is not some amount of positive bias is expected to be present in alpha spec data when compared to ICP-MS data

Considerations

In evaluating which analysis is appropriate for samples from the project neither method is inherently better than the other The following factors should be considered

1 What is being measured ICP-MS measures mass directly and data can be mathematically converted to activity Alpha spec measures activity directly and can be mathematically converted to mass

4 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

2 What are the objectives of the data in terms of measuring total versus environmentally available The ICP-MS method specified in the PRP QAPP measures uranium concentrations in the same manner that EPA has measured other metals concentrations Should uranium mass be measured in a typical way that will be used for measuring other metals of concern at the site (eg copper beryllium arsenic etc)

3 If ICP-MS is more cost effective that alpha spec should the option of using preparation method SW846-3052 be considered This should yield total uranium data similar to alpha spec analysis while still using the efficiencies of ICP-MS analysis

4 What units will the cleanup level for the site be in If its in mass units then ICPshyMS analysis may be more appropriate If its in activity units then alpha spec analysis may be warranted

References

Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts Technical Memorandum from Heath Downey to Bruce Thompson de maxim is inc September 26 2005 (MACTEC 2005)

Test Methods for Evaluating Solid Waste PhysicalChemical Methods SW-846 Third Edition with updates EPA 1996

Ambient Levels of Metals in New Jersey Soils Environmental Assessment and Risk Analysis Element Research Project Summary Division of Science Research and Technology State of New Jersey Department of Environmental Protection May2003

Ground Water Issue Behavior of Metals in Soil EPA Superfund Technology Support Center for Ground Water EPA540S-92018 October 1993

Certificate of Analysis Standard Reference Material 2711 Montana Soil National Institute of Standards amp Technology July 18 2003

bull bull

bull bull bullbull bull

10000

MampE SoilSediment Results Total Uranium mgKg

bull

bull - ~-~-----~~---~~-~--~~1000

en aE

I I bull Total a u

I -

bull bull 100 1shy -middot- -middot bullbull bull

Ybull bull

10 100 1000 10000

Alpha Spec

Attachm

ent 1

10 -

100000

TOTAL URANIUM BY ICP-MS mgKg

bull

bull 10000

I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 5: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

de maximis inc

EPA metals digestion procedures for all analyses (whether ICP ICPMS cold vapor) do not ldquounder-leachrdquo samples rather digestion procedures are designed to evaluate ldquoenvironmentally availablerdquo concentrations As such uranium and thorium would be analyzed as for other potential metal contaminants at the site (Cu Be Pb Hg etc)

While it is in fact possible that some disposed uranium could resist the hot nitric digestion it would also certainly resist leaching by rainwater infiltration or groundwater flushing and be less biologically available if ingested

To assess laboratory performance of the QAPP-approved methods for the analysis of uranium and thorium for the project performance evaluation (PE) samples have been submitted to the laboratory As part of the Phase IA and Phase 1B investigation programs blind PE samples (quality control standard that is of a composition and concentration not known to the laboratory) were submitted to the laboratory for analysis Results of the analyses and comparison to vendor spike concentrations are provided below

Investigation Analysis Parameter Laboratory Result Certified Value (Vendor Spike)

QC PALs

Phase 1A ICPMS Uranium (total)

32 26 24 mgkg1 22 mgkg NA

Phase 1A ICPMS Thorium 229 159 162 110 mgkg1

131 mgkg NA

Phase 1A Alpha Spec

Uraniumshy238

313 pCig 339 pCig NA

Phase 1A Alpha Spec

Uraniumshy234

34 pCig 339 pCig NA

Phase 1A Alpha Spec

Uraniumshy235

0252 U pCig2 016 pCig NA

Phase 1A Alpha Spec

Thorium-230 261 pCig 134 pCig NA

Phase 1B Alpha Spec

Uranium (total)

1184 pCiL 117 pCiL 852 ndash 138 pCiL

Phase 1B ICPMS Uranium 95 mgkg 126 mgkg 834 ndash 144 mgkg

Phase 1B ICPMS Uranium 606 ugL 635 ugL 525 ndash 745 ugL

1 = results provided for multiple runs performed by the laboratory 2 = not detected laboratory minimum detectable activity is greater than amount present in PE sample NA = not available

As seen in this table the laboratory provided acceptable results (ie results generally showed good agreement to vendor certified values and results were reported within QC Performance Acceptance Limits of the vendor result (when applicable) as determined by peer laboratory measurements) indicating the QAPP methods used for uranium and thorium analysis are capable of providing usable results for the RIFS Vendor certified value is a ldquomade tordquo value as determined by volumetric andor gravimetric measurements of National Institute of Standards and Technology (NIST) Standard

5

de maximis inc

Reference Materials or NIST traceable reference material andor analytically verified by the vendor Vendor Performance Acceptance Limits are based on actual historical data collected in the vendors proficiency testing program The limits reflect any inherent biases in the methods used to establish the limits and closely approximate a 95 confidence interval of the performance that experienced laboratories should achieve using accepted environmental methods

An additional statistical evaluation of the analysis of uranium by ICPMS and alpha spectroscopy was performed using linear regression Linear regression mathematically approximates the relationship between two variables and provides an equation to predict a dependent variable from an independent variable Since ICPMS was selected as the primary analytical method for uranium is was assigned as the independent variable (x) and alpha spectroscopy was assigned as the dependent variable (y)

Furthermore the relationship between these two methods was anticipated to be linear such that y = x ideally so the y-intercept was set to equal zero for linear regression During the Fall 2004 RI phase of the investigation 172 soil and sediment samples were collected and analyzed by both analytical methods Of these samples 3 were determined to be from the super-depleted subset by ICPMS and were excluded from linear regression as the results of these samples are questionable

The results of linear regression for this data set are an estimated model of y (alpha spectroscopy) = 10604 x (ICPMS) with an R2 value of 09049 R2 measures the strength of the linear regression model with a value of 1 indicating a perfect linear fit so the results indicate a very strong fit In addition the 95 confidence interval for the slope ranges from 10080 to 11128 which indicates that linear regression modeled alpha spectroscopy results are generally 6 higher than ICPMS results with a range of 1 to 11 (95 confidence) The results of the linear regression analysis are shown in the attached Figure 2 (note that the scale is logarithmic in order to allow better display of all the data points across the range of results) From Figure 2 it is evident that there is substantial scatter in the low concentration range (less than 10 mgkg) and if these data points are removed from the linear regression evaluation there is no significant change in the results since the y-intercept was set to equal zero

These linear regression modeling results provide strong evidence that when the entire population of uranium analysis results for both ICPMS and alpha spectroscopy are considered there is good agreement between the two methods The relative difference is not significant when compared to the expected accuracy of the analytical techniques (+25)

In response to the comment regarding use of alpha spectroscopy values rather than ICPMS values to make remedial decisions it is clear that the commenter is not sufficiently familiar with the CERCLA investigation and remediation process The purpose of the remedial investigation (RI) is to collect data to evaluate the potential risks to human health and the environment associated with chemicals of potential concerns (COPCs) at the Site The data from the two methods show good overall correlation but as indicated there is a discrepancy if only the 1-50 ppm U-238 results are considered

6

de maximis inc

It is important to understand how the RI data will be evaluated and interpreted moving forward in the CERCLA process with respect to the concern that some areas might be missed for remediation The RI data set will be used to derive human exposure point concentrations (EPCs) for COPCs within the various media and exposure areas at the Site EPCs will be derived in accordance with EPA guidance which generally use 95 upper confidence level concentrations for soil analytes For uranium this will be performed using the ICPMS data set As a component of the risk assessment uncertainties associated with the analytical data are evaluated evaluation of alpha spectroscopy outliers as well as evaluation of the effect of uranium enrichment (eg ranging from 02 to 07) on risk estimates will be part of this process

If based on this process a given exposure areamedia has risks that EPA deems to be unacceptable (generally gt 10-4 cancer risk or non-cancer hazard index gt1 based on the cumulative risk from all COPCs in a given mediumexposure area) then preliminary remediation goals (PRGs) will be developed for the COPCs that contribute most substantially to the total risk In turn the PRGs will then be used to evaluate areas and volumes for which remedial alternatives will be considered in the Feasibility Study (FS) Areas for remediation will consider a variety of factors when determining the overall dimensions and they typically will be drawn to include RI data points that are equal to or greater than the PRGs

EPA will then select final remediation goals in the Record of Decision (ROD) The Remedial DesignRemedial Action (RDRA) process occurs after the ROD During the RD areas and volumes for remediation will be revisited based on the final remediation goals typically resulting in identification of a minimum target areavolume that will be remediated As part of the RD process a Demonstration of Compliance Plan will be developed This document sets forth how the contractor will demonstrate that all targeted media has been successfully remediated For a site like NMI we would expect that different instrumentation analytical methods may be used on Site to evaluate the effectiveness of remediation during the RA (eg gamma spectroscopy or XRF which provides faster results than ICPMS or alpha spectroscopy) followed by confirmatory sampling sent to an off-site laboratory Correlation of these instrument results to laboratory analytical methods will be presented at the appropriate time Given this process (CERCLA) and the multiple reviews by EPA State of Massachusetts agencies and public groups it is very unlikely that any areas with unacceptable risk levels will be missed for remediation

With respect to uranium background concentrations the total uranium values calculated using alpha spectroscopy values is correct in the CREW Technical Note (2006) however this background value is based on a theoretical calculation While this calculation is correct de maximis inc (and EPA via MampE split samples) have shown total uranium (reflected in the background evaluation) directly measured by ICPMS is lower due to inherent differences between the two methods and the fact that ICPMS is measuring ldquoenvironmentally availablerdquo concentrations Furthermore the background concentration at the Site is not likely to play a significant role since risk-based remediation levels greater than background are expected to be established for Site clean-up

7

de maximis inc

Lastly the intent for referencing the urine study was to show an independent evaluation that also concludes ICPMS has lower sensitivity for uranium and a greater ability to distinguish between natural depleted and enriched uranium Further it supports the argument that some variability between ICPMS and alpha spectroscopy analysis is not unexpected

In conclusion the ICPMS analysis provides lower results as compared to alpha spectroscopy analysis results however as discussed in this response (and described in detail in MampE uranium evaluation memo) neither method is in error

bull Background uranium concentrations determined from actual ICPMS measurements show environmentally available U background is approximately 13 ppm (EPA MampE-split samples confirm this)

bull lsquoSuper-depletedrsquo points are anomalies and these samples will be recollected and reanalyzed however remaining total U concentrations are usable and ICPMS results generated for the RIFS are reliable

bull Urine analysis referenced in the Technical Memorandum is applicable in that it supports the use of ICPMS for evaluating presence of U235 and calculating enrichments (for future samples once alpha spectroscopy analysis is discontinued)

bull Alpha spectroscopy analysis analyses for the Site were intended to verify enrichment to ensure that no enriched uranium was missed by ICPMS and to confirm signature of uranium and thorium at the site This has been done therefore future alpha spectroscopy analysis is not needed

Response to Section 3 ndash Using enrichment rather than total U for determination of contamination

Using total uranium by mass rather than enrichment to define nature and extent of contamination is being performed at this Site because the determination of enrichment at total uranium concentrations in the background range (eg 075 ndash 25 mgkg) is uncertain particularly for alpha spectroscopy analysis methods where the total propagated uncertainty of measurements is nearly equal to the actual instrument reading This is clearly demonstrated in the scatter observed in data plots provided by CREW (2006) for total uranium in this range de maximis inc concurs that at 5 mgkg a large percentage of the mass should be depleted uranium and for the purposes of defining the nature and extent of contamination based solely on isotopic composition a value of 5 would be inappropriate However the RI is using a value of 21 mgkg as a threshold for the definition of the nature and extent of contamination The nature and extent of uranium like all substances at the Site is being defined by risk-based levels (background levels are only used when background values are higher than risk-based levels) Therefore for total uranium concentrations less than the RI screening level the

8

de maximis inc

source of the uranium (whether depleted or natural uranium) is not relevant For risk assessment purposes all uranium will be considered and the contribution to risk associated with background constituents of concern will be considered after the total risk is calculated

Measurements of environmental data rarely have a normal (Gaussian) distribution and typically are lognormal distributions (Gilbert 1987) EPA guidance indicates that ldquoit is unusual to encounter environmental data sets that are normally distributedrdquo (EPA 2002) and ldquoenvironmental data commonly exhibit frequency distributions that are non-negative and skewed with heavy or long right tailsrdquo (EPA 2000) The composition of uranium (natural or depleted) will be used during risk assessment but it has limited usefulness otherwise However the isotopic composition of uranium will be one of the issues considered during data evaluation to avoid the sort of issues pointed out in the CREW Technical Note

Response to Section 4 - Thorium evaluation

de maximis inc re-iterates that the Technical Memorandum shows thorium-232 and thorium-228 concentrations are roughly equal (and considered within the analytical variability for alpha spectroscopy) and that thorium-230 at the Site is not related to Site activities Furthermore uranium-232 (from recycled uranium) is not present at the site as it is primarily a concern with enriched uranium Even if present it would not be in significant enough quantities to impact thorium-228 concentrations

Refer to previous discussion pertaining to ICPMS versus alpha spectroscopy analysis information Also note that ICPMS instrument sensitivity for thorium is very high due to the high mass to charge ratio (ie it is a good responder) Naturally occurring thorium is 100 Th-232 No data quality issues were observed during data validation and data has been deemed usable

Lastly thorium was used at the Site but at significantly lower amounts when compared to uranium Therefore low concentrations (within background concentration range) of thorium are not unexpected and historical data have demonstrated this

Response to Section 5 - Other radionuclides

There is no way to know the exact pedigree of the depleted uranium processed at the Site or to determine which portion of the depleted uranium might have come from reprocessed fuel rods and therefore contain small amounts of reactor generated radionuclides Studies of depleted uranium have shown that byproduct radionuclides in uranium are present at very low levels In addition dose assessments of these trace contaminants have shown that the radiological dose from these radionuclides is insignificant compared to that from uranium These conclusions were documented in a January 2000 Department of the Army analysis of transuranics and other contaminants in depleted uranium armor (Attachment 3) Of particular relevance is the fact that the Department of the Army analysis took samples of DU billets from Starmet for this study The Army concluded that ldquoThe levels of transuranics elements (TRUs) and fission

9

de maximis inc

products (FP) are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is saferdquo If the TRUs and FP contribute less than a one percent increase in dose for DU metal then they would contribute an even more negligible dose at the 10s to 1000s of mgkg of DU present in NMI Site media With respect to the HTDR analysis during the RI historical data indicated that the holding basin and former drain outfall regions of the bog were the most likely areas to have significant concentrations of uranium therefore the RI work plan identified locations to be sampled in these areas (FSP 2004) Further the approach outlined in the RI was to evaluate for HTDR in samples with elevated concentrations of uranium not necessarily the highest concentrations There were other AOIs where uranium was detected in elevated concentrations but no sample was collected for HTDR However since the holding basin was in use for the vast majority of time that uranium was actively processed at the site and no HTDR were detected in the six samples from this area it is unlikely that HTDR would be identified in other areas of the site

Samples for HTDRs were collected according to the RI (FSP 2004) such that 3 of the samples from AOI 1 were above the water table and 3 of the samples were from below the water table one sample was collected from each of the AOI 6 former drain outfall areas of the bog and one sample collected from the buried drum excavation (AOI 2) These HTDR samples are representative for the areas investigated (AOI 1 AOI 2 and AOI 6) and of the site based on the distribution of total uranium results described below

bull Total uranium in the HTDR sample locations from AOI 1 ranged from 38 to 511 mgkg while total uranium for all samples in AOI 1 ranged from 11 to 672 mgkg Only 25 samples from all samples in AOI 1 had total uranium concentrations greater than 100 mgkg and 4 of the 6 HTDR samples were from this subset

bull Total uranium in the HTDR sample locations from AOI 6 were 25 mgkg at former drain outfall 1 and 4550 mgkg at drain outfall 2 while total uranium for all samples in AOI 6 ranged from 16 to 4550 mgkg Only 10 samples from all samples in AOI 6 had total uranium concentrations greater than 100 mgkg and one HTDR sample was from the maximum total uranium sample location

bull The HTDR sample from AOI 2 drum excavation bottom had a total uranium concentration of 18 mgkg while total uranium for all excavation sides and bottom ranged from 14 to 49 mgkg

Only a few samples have uranium concentrations greater than 1000 mgkg and 6 of the 8 are from samples deemed as lsquosuper-depletedrsquo These 6 locations will be re-sampled as part of Phase 1C since the ICPMS analysis results are considered an anomaly Sporadic low detections of HTDRs do not necessarily indicate the presence of HTDRs as false positive results are common for these analyses Although some HTDRs have been detected in the past they are sporadic (not all samples have detections) and do not indicate the presence of HTDRs at significant levels

10

de maximis inc

It is also important to note that the historical data provided in the RI (PSOP 2005) was used to assist in design of the RI (FSP 2004) Data were compiled from many sources of information and represent samples collected analyzed and reported using various sample collection procedures laboratory analytical methods and data reporting conventions Data compilation activities involved both manual hand-entries and electronic entries when applicable Based on a review of the electronic database severe data quality limitations were noted including unknown analytical quality and validation levels missing location coordinates inconsistent media codes unknown detection limits and radionuclide uncertainty measurements inconsistent andor incorrect reporting of radionuclides using mass-based and activity-based units These limitations were considered significant during the scoping of the RIFS Work Plan therefore the historical data were used primarily to evaluate the spatial distribution of parameters and magnitude of detected concentrations The limitations of the historical data were used to provide the basis for identifying data gaps and scoping the RIFS As such the FSP included analysis for HTDRs at the areas of the Site known to be impacted by historical practices

With respect to comments provided on U-232 and U-236 these uranium isotopes are produced during the fission of uranium (as are plutonium americium and the rest of the HTDRs) Since none of these other fission products have been detected at the Site it is not credible that either uranium-232 or uranium-236 would be present

Response to Section 6 - Radium

The historical radium in groundwater data is suspect for several reasons including the previous discussion on historical data provided in the RI (PSOP 2005) Groundwater samples were collected from 13 wells at the Site (single sampling event) using bailers as part of this investigation All samples were screened for radioactivity by gross alpha beta analysis Only one sample exceeded the action level (gt 10 pCil gross alpha) with results of 50 pCil gross alpha and 55 pCil gross beta and it was analyzed for isotopic uranium Ra-226 and Ra-228 The results were 74 pCil U-234 025 pCil U-235 037 pCil U-238 12 pCil Ra-226 and 7 pCil Ra-228 These values are from Table 10 of the report and it appears that the Ra-226 value was incorrectly transcribed in the CREW Technical Note as 37 pCil The isotopic uranium results do not match those expected as U-234 is elevated and results for U-235 and U-238 are almost equal Therefore the analytical results for this sample are questionable Furthermore all the rest of the ground water samples had gross alpha results of 3 pCil or less indicating that there are no significant amounts of radium in groundwater

In addition a limited literature review was performed to assess the potential for leaching of radium from subsurface solids by nitrate at the NMI Site The literature demonstrates that radium may be found in a handful of aqueous phase complexes The predominant species of radium in aqueous solution is the divalent cation Ra2+ but other complexes such as RaOH+ RaCl+ RaCO3 and RaSO4 may also be found (Beneš et al 1982 Beneš 1984) Radium appears to be the least likely of all the alkali earth metals (beryllium magnesium calcium strontium barium and radium) to form aqueous phase

11

de maximis inc

complexes (Kirby and Salutsky 1964) As such we are unclear as to the origin of Dr Landreacutersquos statement that ldquonitrates are known to influence greatly radium leaching from the soil into the groundwaterrdquo (Landreacute 2006) The salt Ra(NO3)2 is highly soluble (245 g100 g H2O Kirby and Salutsky 1964) but we have not found evidence to suggest that the species Ra(NO3)+ or Ra(NO3)2 exist in aqueous solution Lauria et al (2004) present radionuclide and water chemistry data from groundwater samples taken from the Buena Lagoon in Brazil A plot of Ra-226 and Ra-228 concentrations versus nitrate concentration is presented in Figure 1 These data demonstrate that both Ra-226 and Ra-228 are almost completely uncorrelated with nitrate at this site as evidenced by correlation coefficients (r2) of approximately 005 and 004 respectively Szabo et al (1997) also report that no correlation exists between nitrate and Ra-226 in 17 groundwater samples taken from the Kirkwood-Cohansey aquifer system in New Jersey although Ra-228 appears to be weakly correlated with nitrate in the same data set (r2 = 018)

Figure 1 Correlation between aqueous radium and nitrate (Lauria et al 2004)

r 2 = 005

r 2 = 004

1E-10 1E-09 1E-08 1E-07 1E-06 1E-05 1E-04

0 5 10 15 20 25

Nitrate (mgL)

Ra

(microg

L)

Ra-226 Ra-228 Linear (Ra-226) Linear (Ra-228)

The mobility of radium in the subsurface appears to be controlled primarily by sorption and precipitation (Lauria et al 2004) High values of aqueous phase radium have been observed in groundwater samples with either increased salinity or reduced pH (Kraemer and Reed 1984 Lauria and Godoy 2002) Demir (1989) demonstrated the increased mobility of radium in acidic environments due to the reduced sorptive capacity of sediments at low pH Figure 2 is a plot of radium concentration versus sample pH from the same data set (Lauria et al 2004)

12

de maximis inc

Figure 2 Correlation between aqueous radium and pH (Lauria et al 2004) 226Ra = -4E-06pH + 3E-05

r2 = 058

228Ra = -5E-08pH + 4E-07 r2 = 032

1E-10 1E-09 1E-08 1E-07 1E-06 1E-05 1E-04

3 4 5 6 7

pH

Ra

(microg

L)

Ra-226 Ra-228 Linear (Ra-226) Linear (Ra-228)

Here a stronger correlation between Ra-226 concentration and pH exists (r2 = 058) along with an observable trend of decreasing aqueous phase radium with increasing pH Szabo et al (1997) also report a negative correlation (r2 = 030) between Ra-228 and pH but no significant correlation between Ra-226 and pH Divalent cations such as calcium and magnesium compete with radium for a limited number of surface complexation and adsorption sites (Szabo et al 1997) As lime (a mixture potentially containing calcium oxide calcium hydroxide and calcium carbonate) was co-disposed with uranium and nitric acid at the Holding Basin nitrate may serve as a site-specific tracer for conditions that may lead to the release of adsorbed or surface-complexed radium from the solid phase although no comparable calcium ldquoplumerdquo has been defined unlike nitrate in groundwater Therefore we suggest that Dr Landreacutersquos statement could be qualified by stating that low pH values are known to influence the leaching of radium from aquifer solids However as discussed above we do not see a compelling reason to add radium to the analytical list at this time

Please contact me if you have any questions

Sincerely

Bruce Thompson

cc Respondents Bennie Underwood PE de maximis inc inc Nelson Walter PE MACTEC Inc Peter Zeeb PhD PG GeoSyntec Consultants Inc

13

de maximis inc

Figure 1 ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

Figure 2 Comparison of Alpha Spectroscopy and ICP-MS Uranium Analysis

Attachment 1 Metcalf amp Eddy Memorandum ndash Uranium Analysis comparison of alpha spectrometry and ICP-MS results

Attachment 2 QAPP SOPs for Alpha Spectroscopy and ICPMS Sample Preparation

Attachment 3 Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor Department of the Army January 19 2000

References

Beneš P Migration of radium in the terrestrial hydrosphere The behavior of radium in waterways and aquifers Technical Reports Series No 310 International Atomic Energy Agency IAEA-TECDOC-301 Vienna Austrai pp 119-173 1984

Beneš P M Obdržaacutelek M Čejchanovaacute The physicochemical forms of tracers of radium in aqueous solutions containing chlorides sulfates and carbonates Radiochem Radioanal Letters 50(4) pp 227-242 1982

CREW 2006 ldquoComments on Radiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Technical Note No 2006-01 January 6 2006

Demir I Temporal variations and sources of Pb Co Cu Ni Fe and Mo in shallow ground water of the McDonalds Branch basin Lebanon State Forest Burlington County New Jersey New Jersey Geological Survey Report 20 19 p 1989

EPA 2000 Guidance for Data Quality Assessment Practical Methods for Data Analysis EPA QAG-9 July 2000

EPA 2002 Guidance for Comparing Background and Chemical Concentrations in Soil for CERCLA Site EPA 540-R-01-003 September 2002

FSP 2004 Field Sampling Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Gilbert 1989 Statistical Methods for Environmental Pollution Monitoring Richard O Gilbert 1987

Kirby H W M L Salutsky The Radiochemistry of Radium National Academy of Sciences ndash National Research Council NAS-NS 3057 205 p 1964

14

de maximis inc

Kramer T F and D F Reid The occurrence and behavior of radium in saline formation water of the US Gulf Coast Region Isotope Geoscience 2 pp 153shy174 1984

Landreacute V Comments on ldquoRadiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Nuclear Metals Superfund Site CREW Technical Note 2006-01 January 6 2006

Lauria D C and J M O Godoy Abnormal high natural radium concentration in surface waters Journal of Environmental Radioactivity 61(2) pp 27-37 2002

Lauria D C R M R Almeida O Sracek Behavior of radium thorium and uranium in groundwater near the Buena Lagoon in the Coastal Zone of the State of Rio de Janeiro Brazil Environmental Geology 47 pp 11-19 200

MACTEC 2005 Radiological Review of Fall 2004 RI Analytical Data Nuclear Metals Superfund Site Concord Massachusetts September 26 2004

MampE 2005 Memorandum from Andrew Schkuta to Ed Conroy Subject Uranium Analysis comparison of alpha spectrometry and ICPMS results December 15 2005

PSOP Project Summary and Operations Plan Nuclear Metals Superfund Site Concord Massachusetts April 15 2005

QAPP 2004 Quality Assurance Project Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Szabo Z D E Rice C L MacLeod and T H Barringer Relation of distribution of radium nitrate and pesticides to agricultural land use and depth Kirkwood-Cohansey aquifer system New Jersey Coastal Plain 1990-1991 US Geological Survey Water-Resources Investigations Report 96-4165A 119 p 1997

15

bull

bull bull

bull bull

bull bull bull

bull bull

bull bull

bull

500

1500

Figure 1

Con

cent

ratio

n (m

gkg

)

ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

1000

0

0 5 10 15 20 25 30 35 40

Samples

Figure 2

Alp

ha S

pect

osco

py (m

gkg

)

Fall 2004 Total Uranium No Super-Depleted

1000

100

10

1

y = 10604x R2 = 09049

1 10 100 1000

ICPMS (mgkg)

Fall 2004 Super-Depleted Linear Regression

Attachment 1Connolly Jayme

From Taylor Mel issagepam ail epa gov Sent Wednesday December 21 2005 932AM To brucetdemaximiscom jhuntdemaximiscom Subject Fw Uranium Analysis Issues

Rad-versus-ICPMS RADvsiCPMS_soilse MEvsPRP_ICPMSP memodoc (158 dPDF (29 KB) OF (25 KB)

brucejohn here is mampes take on the rns vs alpha spec analysis discrepancy

Forwarded by MelissaG TaylorR1USEPAUS on 12212005 0930 AM

Conroy Ed ltEdConroyregrn-ea ecomcomgt To

MelissaG TaylorR1USEPAUSregEPA 12152005 0416 cc PM

Subject FW Uranium Analysis Issues

Melissa

Heres Andys take on the alpha spec vs ICPMS debate along with a couple of graphs showing correlation between the two methods and MampE split data vs PRP data

Please call me on the cell phone tomorrow to discuss

Ed

Edward A Conroy PE

Project Manager

METCALF amp EDDYJAECOM

7812246880 ph

6172832942 cell

1

ed conroym-e aecom com Attachment 1

-----Original Message----shyFrom Schkuta Andrew Sent Thursday December 15 2005 358 PM To Conroy Ed Subject Uranium Analysis Issues

Ed

Please find attached a memorandum that summarizes my review of why ICP-MS and alpha spec data may not necessarily agree

ltltRad-versus-ICPMS memodocgtgt

In addition I have attached 2 preliminary charts that have the following draft information

ltltRADvsiCPMS soilsedPDFgtgt

The RADvsiCPMS soilsed chart is a draft plot of MampEs alpha spec versus ICP-MS results when data was available for both samples It shows the same general trend as the PRP data with fairly good agreement between the two analyses There appears to be a slight positive bias for alpha spec data at low concentrations but this is expected as discussed in my memo

ltltMEvsPRP ICPMSPDFgtgt

The MEvsPRP ICPMS is a draft plot of MampE vs the PRPs total uranium results as measured by ICP-MS -With the exception of a few outliers which will be further investigated there appears to be good agreement between MampEs and the PRPs data As a note the median percent difference between MampEs and the PRPs data was 1

Andrew Schkuta CHMM

Technical Specialist - Chemistry

METCALF amp EDDYJAECOM

701 Edgewater Drive

Wakefield MA 01880

(781) 224-6353

(781) 245-6293 fax

2

(See attached file Rad-versus-ICPMS Attachment 1memodoc) (See attached file RADvsiCPMS soilsedPDF) (See attached file MEvsPRP ICPMSPDF)

3

Attachment 1

Metcalf amp Eddy Inc

METCALFampEDDY

Memorandum

Date December 15 2005

To Ed Conroy

From Andrew Schkuta

Subject Uranium Analysis comparison of alpha spectrometry and ICP-MS results

File WA151 Nuclear Metals Distribution

Summary This memorandum discusses the differences between uranium analysis conducted by two different methods alpha spectroscopy (alpha spec) and inductively coupled plasma shymass spectrometry (ICP-MS) In summary the alpha spec data is expected to be biased high compared to the ICP-MS data but this is not indicative that either method is being performed incorrectly

In the September 25 2005 technical memorandum Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts (MACTEC 2005) data was presented comparing alpha spec and ICP-MS analysis for uranium analysis and it was proposed that future uranium analysis be performed by ICP-MS only For samples analyzed that were collected through Spring 2005 all samples were analyzed by ICP-MS with a subset (20) analyzed by alpha spec It was noted that there appeared to be a positive bias in the alpha spec data compared to the ICP-MS data and some concern has been voiced if analysis by ICP-MS only would meet project objectives This memorandum provides background on why the two analytical methods may yield differing results and discusses how they may impact project quality objectives

There are two primary reasons why alpha spec is expected to yield results that may be positively biased

1 Sample Preparation The sample preparation method for alpha spec is more rigorous The preparation method in alpha spec breaks down the silica mineral structure of the sample By design the ICP-MS sample preparation does not break down the mineral structure and is intended to remove all metals that could become environmentally available

2 Sample Quanitation The results for alpha spec are adjusted for analyte loss by correcting the result based on the amount of tracer recovered (yield) ICP-MS data is absolute ie data is not corrected for loss during sample preparation

2 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

These two reasons are discussed in further detail as follows

Sample Preparation

Radionuclide analysis traditionally uses a sample preparation method that includes hydrofluoric acid (HF) to break down the sample so that all of the uranium present in the sample is put into solution On the other hand standard EPA methods for metals analysis including ICP-MS do not measure the total metals present in a sample For example method SW846-3050 (EPA 1996) (as specified by the PRP in their approved QAPP for metals analysis by ICP-MS) states

This method is not a total digestion technique for most samples It is a very strong acid digestion that will dissolve almost all elements that could become environmentally available By design elements bound in silicate structures are not normally dissolved by this procedure as they are not usually mobile in the environment If absolute total digestion is required use Method 3052

As noted by the method a SW846 method is available to do a total digestion (equivalent to the alpha spec method) as method 3052 However method 3052 (EPA 1996) states (emphasis added)

Note This technique is not appropriate for regulatory applications that require the use of leachate preparations (ie Method 3050 Method 3051 Method 1311 Method 1312 Method 1310 Method 1320 Method 1330 Method 3031 Method 3040) This method is appropriate for those applications requiring a total decomposition for research purposes (ie~ geological studiesr mass balancesr analysis of Standard Reference Materials) or in response to a regulation that requires total sample decomposition

Even though the ICP-MS preparation method does not yield total results it still may conservatively measure the amount of metals that are environmentally available In the Research Project Summary Ambient Levels of Metals in New Jersey Soils prepared by the New Jersey Department of Environmental Protection (NJDEP 2003) where samples were prepared using SW846-3050 procedures it states

Metal concentrations reported in this summary were acid-extractable metals not necessarily total metal concentrations The extraction method (USEPA Method 3050) is a vigorous extraction method designed to remove all metals that could become environmentally available In practice the extraction method likely overestimates metals that could become available since the sample is refluxed with both concentrated nitric acid and hydrogen peroxide However by design the method will not extract chemicals incorporated in silica minerals as they are usually not mobile in the environment Thus concentrations reported in these studies may be lower than those indicated from analyses using methods designed to measure total metal concentrations such as x-ray fluorescence methods

Although references to uranium were not located it is expected that uranium will have some of the same type of mobility characteristics as other metals in the soil matrix The paper Ground Water Issue Behavior of Metals in Soil published by EPAs Superfund Technology Support Center for Ground Water (EPA 1992) describes how metals are distributed within soil The paper describes how metals are partitioned in various pools and how a rigorous extraction using HF is required to release metals present in the structure of primary or secondary minerals It is

3 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

expected that native metals (ie background concentrations) can be present in these structures The paper indicates that for four metals examined (cadmium copper zinc and nickel) the amount of metals that was bound in the mineral structure ranges from approximately 15-60 It would be expected that naturally occurring uranium would have similar characteristics

The National Institutes of Standards amp Technology (NIST) recognizes the differences between total digestion and method SW846-3050 and has issued Addendum to Certificates for the Certificate of Analysis (NIST 2003) for some Standard Reference Materials (SRMs) The Addendum to Certificates for SRM 2711 a Montana agricultural soil indicates that trace metals were measured at approximately 40 to 90 of their certified value when analyzed using SW846-3050 as the preparation method Even though uranium is not listed as a target analyte it is reasonable to expect that is would behave in a similar fashion

Sample Quanitation

The preparation method for alpha spec analysis includes addition of a tracer (U-232) to sample and is carried through the entire analytical scheme The amount of tracer is measured during analysis and the results for target compounds (the other uranium isotopes) are corrected for the amount of tracer recovery (yield) For example if amount of tracer measured is 50 of what was initially added to the sample the sample results are all corrected by dividing by 050 In terms of method quality assurancequality control (QAQC) the PRP QAPP indicates that tracer recovery in the range of 15-125 is considered acceptable and the analytical system is considered in control In other words up to 85 of the compound can be lost during sample preparation and analysis and the data is considered meeting all QAQC objectives This method of analysis is the equivalent to the technique of isotope dilution which is used for some non-radiological environmental analyses (eg EPA Method 1625 Semivolatile Organic Compounds by Isotope Dilution GCMS) Analyses using this technique although considered the most accurate are not widely employed due to the complexity and expense of using the isotope dilution technique

ICP-MS analysis measures the concentration of analyte with no correction for potential losses However QAQC procedures are in place to make sure the losses are not considered significant In addition to laboratory control samples matrix spike analysis measures the recovery for the target analyte that was over-spiked into the sample The PRP QAPP states that matrix spike recovery should be within the range of 75-125 Therefore up to 25 of the analyte could be lost and the method has still met the QAQC objectives If recovery is less than this then validation procedures acknowledge that the data may be biased low and sample results are flagged as estimated

Therefore because the alpha spec data is corrected for analyte loss during the analytical process while ICP-MS data is not some amount of positive bias is expected to be present in alpha spec data when compared to ICP-MS data

Considerations

In evaluating which analysis is appropriate for samples from the project neither method is inherently better than the other The following factors should be considered

1 What is being measured ICP-MS measures mass directly and data can be mathematically converted to activity Alpha spec measures activity directly and can be mathematically converted to mass

4 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

2 What are the objectives of the data in terms of measuring total versus environmentally available The ICP-MS method specified in the PRP QAPP measures uranium concentrations in the same manner that EPA has measured other metals concentrations Should uranium mass be measured in a typical way that will be used for measuring other metals of concern at the site (eg copper beryllium arsenic etc)

3 If ICP-MS is more cost effective that alpha spec should the option of using preparation method SW846-3052 be considered This should yield total uranium data similar to alpha spec analysis while still using the efficiencies of ICP-MS analysis

4 What units will the cleanup level for the site be in If its in mass units then ICPshyMS analysis may be more appropriate If its in activity units then alpha spec analysis may be warranted

References

Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts Technical Memorandum from Heath Downey to Bruce Thompson de maxim is inc September 26 2005 (MACTEC 2005)

Test Methods for Evaluating Solid Waste PhysicalChemical Methods SW-846 Third Edition with updates EPA 1996

Ambient Levels of Metals in New Jersey Soils Environmental Assessment and Risk Analysis Element Research Project Summary Division of Science Research and Technology State of New Jersey Department of Environmental Protection May2003

Ground Water Issue Behavior of Metals in Soil EPA Superfund Technology Support Center for Ground Water EPA540S-92018 October 1993

Certificate of Analysis Standard Reference Material 2711 Montana Soil National Institute of Standards amp Technology July 18 2003

bull bull

bull bull bullbull bull

10000

MampE SoilSediment Results Total Uranium mgKg

bull

bull - ~-~-----~~---~~-~--~~1000

en aE

I I bull Total a u

I -

bull bull 100 1shy -middot- -middot bullbull bull

Ybull bull

10 100 1000 10000

Alpha Spec

Attachm

ent 1

10 -

100000

TOTAL URANIUM BY ICP-MS mgKg

bull

bull 10000

I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 6: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

de maximis inc

Reference Materials or NIST traceable reference material andor analytically verified by the vendor Vendor Performance Acceptance Limits are based on actual historical data collected in the vendors proficiency testing program The limits reflect any inherent biases in the methods used to establish the limits and closely approximate a 95 confidence interval of the performance that experienced laboratories should achieve using accepted environmental methods

An additional statistical evaluation of the analysis of uranium by ICPMS and alpha spectroscopy was performed using linear regression Linear regression mathematically approximates the relationship between two variables and provides an equation to predict a dependent variable from an independent variable Since ICPMS was selected as the primary analytical method for uranium is was assigned as the independent variable (x) and alpha spectroscopy was assigned as the dependent variable (y)

Furthermore the relationship between these two methods was anticipated to be linear such that y = x ideally so the y-intercept was set to equal zero for linear regression During the Fall 2004 RI phase of the investigation 172 soil and sediment samples were collected and analyzed by both analytical methods Of these samples 3 were determined to be from the super-depleted subset by ICPMS and were excluded from linear regression as the results of these samples are questionable

The results of linear regression for this data set are an estimated model of y (alpha spectroscopy) = 10604 x (ICPMS) with an R2 value of 09049 R2 measures the strength of the linear regression model with a value of 1 indicating a perfect linear fit so the results indicate a very strong fit In addition the 95 confidence interval for the slope ranges from 10080 to 11128 which indicates that linear regression modeled alpha spectroscopy results are generally 6 higher than ICPMS results with a range of 1 to 11 (95 confidence) The results of the linear regression analysis are shown in the attached Figure 2 (note that the scale is logarithmic in order to allow better display of all the data points across the range of results) From Figure 2 it is evident that there is substantial scatter in the low concentration range (less than 10 mgkg) and if these data points are removed from the linear regression evaluation there is no significant change in the results since the y-intercept was set to equal zero

These linear regression modeling results provide strong evidence that when the entire population of uranium analysis results for both ICPMS and alpha spectroscopy are considered there is good agreement between the two methods The relative difference is not significant when compared to the expected accuracy of the analytical techniques (+25)

In response to the comment regarding use of alpha spectroscopy values rather than ICPMS values to make remedial decisions it is clear that the commenter is not sufficiently familiar with the CERCLA investigation and remediation process The purpose of the remedial investigation (RI) is to collect data to evaluate the potential risks to human health and the environment associated with chemicals of potential concerns (COPCs) at the Site The data from the two methods show good overall correlation but as indicated there is a discrepancy if only the 1-50 ppm U-238 results are considered

6

de maximis inc

It is important to understand how the RI data will be evaluated and interpreted moving forward in the CERCLA process with respect to the concern that some areas might be missed for remediation The RI data set will be used to derive human exposure point concentrations (EPCs) for COPCs within the various media and exposure areas at the Site EPCs will be derived in accordance with EPA guidance which generally use 95 upper confidence level concentrations for soil analytes For uranium this will be performed using the ICPMS data set As a component of the risk assessment uncertainties associated with the analytical data are evaluated evaluation of alpha spectroscopy outliers as well as evaluation of the effect of uranium enrichment (eg ranging from 02 to 07) on risk estimates will be part of this process

If based on this process a given exposure areamedia has risks that EPA deems to be unacceptable (generally gt 10-4 cancer risk or non-cancer hazard index gt1 based on the cumulative risk from all COPCs in a given mediumexposure area) then preliminary remediation goals (PRGs) will be developed for the COPCs that contribute most substantially to the total risk In turn the PRGs will then be used to evaluate areas and volumes for which remedial alternatives will be considered in the Feasibility Study (FS) Areas for remediation will consider a variety of factors when determining the overall dimensions and they typically will be drawn to include RI data points that are equal to or greater than the PRGs

EPA will then select final remediation goals in the Record of Decision (ROD) The Remedial DesignRemedial Action (RDRA) process occurs after the ROD During the RD areas and volumes for remediation will be revisited based on the final remediation goals typically resulting in identification of a minimum target areavolume that will be remediated As part of the RD process a Demonstration of Compliance Plan will be developed This document sets forth how the contractor will demonstrate that all targeted media has been successfully remediated For a site like NMI we would expect that different instrumentation analytical methods may be used on Site to evaluate the effectiveness of remediation during the RA (eg gamma spectroscopy or XRF which provides faster results than ICPMS or alpha spectroscopy) followed by confirmatory sampling sent to an off-site laboratory Correlation of these instrument results to laboratory analytical methods will be presented at the appropriate time Given this process (CERCLA) and the multiple reviews by EPA State of Massachusetts agencies and public groups it is very unlikely that any areas with unacceptable risk levels will be missed for remediation

With respect to uranium background concentrations the total uranium values calculated using alpha spectroscopy values is correct in the CREW Technical Note (2006) however this background value is based on a theoretical calculation While this calculation is correct de maximis inc (and EPA via MampE split samples) have shown total uranium (reflected in the background evaluation) directly measured by ICPMS is lower due to inherent differences between the two methods and the fact that ICPMS is measuring ldquoenvironmentally availablerdquo concentrations Furthermore the background concentration at the Site is not likely to play a significant role since risk-based remediation levels greater than background are expected to be established for Site clean-up

7

de maximis inc

Lastly the intent for referencing the urine study was to show an independent evaluation that also concludes ICPMS has lower sensitivity for uranium and a greater ability to distinguish between natural depleted and enriched uranium Further it supports the argument that some variability between ICPMS and alpha spectroscopy analysis is not unexpected

In conclusion the ICPMS analysis provides lower results as compared to alpha spectroscopy analysis results however as discussed in this response (and described in detail in MampE uranium evaluation memo) neither method is in error

bull Background uranium concentrations determined from actual ICPMS measurements show environmentally available U background is approximately 13 ppm (EPA MampE-split samples confirm this)

bull lsquoSuper-depletedrsquo points are anomalies and these samples will be recollected and reanalyzed however remaining total U concentrations are usable and ICPMS results generated for the RIFS are reliable

bull Urine analysis referenced in the Technical Memorandum is applicable in that it supports the use of ICPMS for evaluating presence of U235 and calculating enrichments (for future samples once alpha spectroscopy analysis is discontinued)

bull Alpha spectroscopy analysis analyses for the Site were intended to verify enrichment to ensure that no enriched uranium was missed by ICPMS and to confirm signature of uranium and thorium at the site This has been done therefore future alpha spectroscopy analysis is not needed

Response to Section 3 ndash Using enrichment rather than total U for determination of contamination

Using total uranium by mass rather than enrichment to define nature and extent of contamination is being performed at this Site because the determination of enrichment at total uranium concentrations in the background range (eg 075 ndash 25 mgkg) is uncertain particularly for alpha spectroscopy analysis methods where the total propagated uncertainty of measurements is nearly equal to the actual instrument reading This is clearly demonstrated in the scatter observed in data plots provided by CREW (2006) for total uranium in this range de maximis inc concurs that at 5 mgkg a large percentage of the mass should be depleted uranium and for the purposes of defining the nature and extent of contamination based solely on isotopic composition a value of 5 would be inappropriate However the RI is using a value of 21 mgkg as a threshold for the definition of the nature and extent of contamination The nature and extent of uranium like all substances at the Site is being defined by risk-based levels (background levels are only used when background values are higher than risk-based levels) Therefore for total uranium concentrations less than the RI screening level the

8

de maximis inc

source of the uranium (whether depleted or natural uranium) is not relevant For risk assessment purposes all uranium will be considered and the contribution to risk associated with background constituents of concern will be considered after the total risk is calculated

Measurements of environmental data rarely have a normal (Gaussian) distribution and typically are lognormal distributions (Gilbert 1987) EPA guidance indicates that ldquoit is unusual to encounter environmental data sets that are normally distributedrdquo (EPA 2002) and ldquoenvironmental data commonly exhibit frequency distributions that are non-negative and skewed with heavy or long right tailsrdquo (EPA 2000) The composition of uranium (natural or depleted) will be used during risk assessment but it has limited usefulness otherwise However the isotopic composition of uranium will be one of the issues considered during data evaluation to avoid the sort of issues pointed out in the CREW Technical Note

Response to Section 4 - Thorium evaluation

de maximis inc re-iterates that the Technical Memorandum shows thorium-232 and thorium-228 concentrations are roughly equal (and considered within the analytical variability for alpha spectroscopy) and that thorium-230 at the Site is not related to Site activities Furthermore uranium-232 (from recycled uranium) is not present at the site as it is primarily a concern with enriched uranium Even if present it would not be in significant enough quantities to impact thorium-228 concentrations

Refer to previous discussion pertaining to ICPMS versus alpha spectroscopy analysis information Also note that ICPMS instrument sensitivity for thorium is very high due to the high mass to charge ratio (ie it is a good responder) Naturally occurring thorium is 100 Th-232 No data quality issues were observed during data validation and data has been deemed usable

Lastly thorium was used at the Site but at significantly lower amounts when compared to uranium Therefore low concentrations (within background concentration range) of thorium are not unexpected and historical data have demonstrated this

Response to Section 5 - Other radionuclides

There is no way to know the exact pedigree of the depleted uranium processed at the Site or to determine which portion of the depleted uranium might have come from reprocessed fuel rods and therefore contain small amounts of reactor generated radionuclides Studies of depleted uranium have shown that byproduct radionuclides in uranium are present at very low levels In addition dose assessments of these trace contaminants have shown that the radiological dose from these radionuclides is insignificant compared to that from uranium These conclusions were documented in a January 2000 Department of the Army analysis of transuranics and other contaminants in depleted uranium armor (Attachment 3) Of particular relevance is the fact that the Department of the Army analysis took samples of DU billets from Starmet for this study The Army concluded that ldquoThe levels of transuranics elements (TRUs) and fission

9

de maximis inc

products (FP) are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is saferdquo If the TRUs and FP contribute less than a one percent increase in dose for DU metal then they would contribute an even more negligible dose at the 10s to 1000s of mgkg of DU present in NMI Site media With respect to the HTDR analysis during the RI historical data indicated that the holding basin and former drain outfall regions of the bog were the most likely areas to have significant concentrations of uranium therefore the RI work plan identified locations to be sampled in these areas (FSP 2004) Further the approach outlined in the RI was to evaluate for HTDR in samples with elevated concentrations of uranium not necessarily the highest concentrations There were other AOIs where uranium was detected in elevated concentrations but no sample was collected for HTDR However since the holding basin was in use for the vast majority of time that uranium was actively processed at the site and no HTDR were detected in the six samples from this area it is unlikely that HTDR would be identified in other areas of the site

Samples for HTDRs were collected according to the RI (FSP 2004) such that 3 of the samples from AOI 1 were above the water table and 3 of the samples were from below the water table one sample was collected from each of the AOI 6 former drain outfall areas of the bog and one sample collected from the buried drum excavation (AOI 2) These HTDR samples are representative for the areas investigated (AOI 1 AOI 2 and AOI 6) and of the site based on the distribution of total uranium results described below

bull Total uranium in the HTDR sample locations from AOI 1 ranged from 38 to 511 mgkg while total uranium for all samples in AOI 1 ranged from 11 to 672 mgkg Only 25 samples from all samples in AOI 1 had total uranium concentrations greater than 100 mgkg and 4 of the 6 HTDR samples were from this subset

bull Total uranium in the HTDR sample locations from AOI 6 were 25 mgkg at former drain outfall 1 and 4550 mgkg at drain outfall 2 while total uranium for all samples in AOI 6 ranged from 16 to 4550 mgkg Only 10 samples from all samples in AOI 6 had total uranium concentrations greater than 100 mgkg and one HTDR sample was from the maximum total uranium sample location

bull The HTDR sample from AOI 2 drum excavation bottom had a total uranium concentration of 18 mgkg while total uranium for all excavation sides and bottom ranged from 14 to 49 mgkg

Only a few samples have uranium concentrations greater than 1000 mgkg and 6 of the 8 are from samples deemed as lsquosuper-depletedrsquo These 6 locations will be re-sampled as part of Phase 1C since the ICPMS analysis results are considered an anomaly Sporadic low detections of HTDRs do not necessarily indicate the presence of HTDRs as false positive results are common for these analyses Although some HTDRs have been detected in the past they are sporadic (not all samples have detections) and do not indicate the presence of HTDRs at significant levels

10

de maximis inc

It is also important to note that the historical data provided in the RI (PSOP 2005) was used to assist in design of the RI (FSP 2004) Data were compiled from many sources of information and represent samples collected analyzed and reported using various sample collection procedures laboratory analytical methods and data reporting conventions Data compilation activities involved both manual hand-entries and electronic entries when applicable Based on a review of the electronic database severe data quality limitations were noted including unknown analytical quality and validation levels missing location coordinates inconsistent media codes unknown detection limits and radionuclide uncertainty measurements inconsistent andor incorrect reporting of radionuclides using mass-based and activity-based units These limitations were considered significant during the scoping of the RIFS Work Plan therefore the historical data were used primarily to evaluate the spatial distribution of parameters and magnitude of detected concentrations The limitations of the historical data were used to provide the basis for identifying data gaps and scoping the RIFS As such the FSP included analysis for HTDRs at the areas of the Site known to be impacted by historical practices

With respect to comments provided on U-232 and U-236 these uranium isotopes are produced during the fission of uranium (as are plutonium americium and the rest of the HTDRs) Since none of these other fission products have been detected at the Site it is not credible that either uranium-232 or uranium-236 would be present

Response to Section 6 - Radium

The historical radium in groundwater data is suspect for several reasons including the previous discussion on historical data provided in the RI (PSOP 2005) Groundwater samples were collected from 13 wells at the Site (single sampling event) using bailers as part of this investigation All samples were screened for radioactivity by gross alpha beta analysis Only one sample exceeded the action level (gt 10 pCil gross alpha) with results of 50 pCil gross alpha and 55 pCil gross beta and it was analyzed for isotopic uranium Ra-226 and Ra-228 The results were 74 pCil U-234 025 pCil U-235 037 pCil U-238 12 pCil Ra-226 and 7 pCil Ra-228 These values are from Table 10 of the report and it appears that the Ra-226 value was incorrectly transcribed in the CREW Technical Note as 37 pCil The isotopic uranium results do not match those expected as U-234 is elevated and results for U-235 and U-238 are almost equal Therefore the analytical results for this sample are questionable Furthermore all the rest of the ground water samples had gross alpha results of 3 pCil or less indicating that there are no significant amounts of radium in groundwater

In addition a limited literature review was performed to assess the potential for leaching of radium from subsurface solids by nitrate at the NMI Site The literature demonstrates that radium may be found in a handful of aqueous phase complexes The predominant species of radium in aqueous solution is the divalent cation Ra2+ but other complexes such as RaOH+ RaCl+ RaCO3 and RaSO4 may also be found (Beneš et al 1982 Beneš 1984) Radium appears to be the least likely of all the alkali earth metals (beryllium magnesium calcium strontium barium and radium) to form aqueous phase

11

de maximis inc

complexes (Kirby and Salutsky 1964) As such we are unclear as to the origin of Dr Landreacutersquos statement that ldquonitrates are known to influence greatly radium leaching from the soil into the groundwaterrdquo (Landreacute 2006) The salt Ra(NO3)2 is highly soluble (245 g100 g H2O Kirby and Salutsky 1964) but we have not found evidence to suggest that the species Ra(NO3)+ or Ra(NO3)2 exist in aqueous solution Lauria et al (2004) present radionuclide and water chemistry data from groundwater samples taken from the Buena Lagoon in Brazil A plot of Ra-226 and Ra-228 concentrations versus nitrate concentration is presented in Figure 1 These data demonstrate that both Ra-226 and Ra-228 are almost completely uncorrelated with nitrate at this site as evidenced by correlation coefficients (r2) of approximately 005 and 004 respectively Szabo et al (1997) also report that no correlation exists between nitrate and Ra-226 in 17 groundwater samples taken from the Kirkwood-Cohansey aquifer system in New Jersey although Ra-228 appears to be weakly correlated with nitrate in the same data set (r2 = 018)

Figure 1 Correlation between aqueous radium and nitrate (Lauria et al 2004)

r 2 = 005

r 2 = 004

1E-10 1E-09 1E-08 1E-07 1E-06 1E-05 1E-04

0 5 10 15 20 25

Nitrate (mgL)

Ra

(microg

L)

Ra-226 Ra-228 Linear (Ra-226) Linear (Ra-228)

The mobility of radium in the subsurface appears to be controlled primarily by sorption and precipitation (Lauria et al 2004) High values of aqueous phase radium have been observed in groundwater samples with either increased salinity or reduced pH (Kraemer and Reed 1984 Lauria and Godoy 2002) Demir (1989) demonstrated the increased mobility of radium in acidic environments due to the reduced sorptive capacity of sediments at low pH Figure 2 is a plot of radium concentration versus sample pH from the same data set (Lauria et al 2004)

12

de maximis inc

Figure 2 Correlation between aqueous radium and pH (Lauria et al 2004) 226Ra = -4E-06pH + 3E-05

r2 = 058

228Ra = -5E-08pH + 4E-07 r2 = 032

1E-10 1E-09 1E-08 1E-07 1E-06 1E-05 1E-04

3 4 5 6 7

pH

Ra

(microg

L)

Ra-226 Ra-228 Linear (Ra-226) Linear (Ra-228)

Here a stronger correlation between Ra-226 concentration and pH exists (r2 = 058) along with an observable trend of decreasing aqueous phase radium with increasing pH Szabo et al (1997) also report a negative correlation (r2 = 030) between Ra-228 and pH but no significant correlation between Ra-226 and pH Divalent cations such as calcium and magnesium compete with radium for a limited number of surface complexation and adsorption sites (Szabo et al 1997) As lime (a mixture potentially containing calcium oxide calcium hydroxide and calcium carbonate) was co-disposed with uranium and nitric acid at the Holding Basin nitrate may serve as a site-specific tracer for conditions that may lead to the release of adsorbed or surface-complexed radium from the solid phase although no comparable calcium ldquoplumerdquo has been defined unlike nitrate in groundwater Therefore we suggest that Dr Landreacutersquos statement could be qualified by stating that low pH values are known to influence the leaching of radium from aquifer solids However as discussed above we do not see a compelling reason to add radium to the analytical list at this time

Please contact me if you have any questions

Sincerely

Bruce Thompson

cc Respondents Bennie Underwood PE de maximis inc inc Nelson Walter PE MACTEC Inc Peter Zeeb PhD PG GeoSyntec Consultants Inc

13

de maximis inc

Figure 1 ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

Figure 2 Comparison of Alpha Spectroscopy and ICP-MS Uranium Analysis

Attachment 1 Metcalf amp Eddy Memorandum ndash Uranium Analysis comparison of alpha spectrometry and ICP-MS results

Attachment 2 QAPP SOPs for Alpha Spectroscopy and ICPMS Sample Preparation

Attachment 3 Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor Department of the Army January 19 2000

References

Beneš P Migration of radium in the terrestrial hydrosphere The behavior of radium in waterways and aquifers Technical Reports Series No 310 International Atomic Energy Agency IAEA-TECDOC-301 Vienna Austrai pp 119-173 1984

Beneš P M Obdržaacutelek M Čejchanovaacute The physicochemical forms of tracers of radium in aqueous solutions containing chlorides sulfates and carbonates Radiochem Radioanal Letters 50(4) pp 227-242 1982

CREW 2006 ldquoComments on Radiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Technical Note No 2006-01 January 6 2006

Demir I Temporal variations and sources of Pb Co Cu Ni Fe and Mo in shallow ground water of the McDonalds Branch basin Lebanon State Forest Burlington County New Jersey New Jersey Geological Survey Report 20 19 p 1989

EPA 2000 Guidance for Data Quality Assessment Practical Methods for Data Analysis EPA QAG-9 July 2000

EPA 2002 Guidance for Comparing Background and Chemical Concentrations in Soil for CERCLA Site EPA 540-R-01-003 September 2002

FSP 2004 Field Sampling Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Gilbert 1989 Statistical Methods for Environmental Pollution Monitoring Richard O Gilbert 1987

Kirby H W M L Salutsky The Radiochemistry of Radium National Academy of Sciences ndash National Research Council NAS-NS 3057 205 p 1964

14

de maximis inc

Kramer T F and D F Reid The occurrence and behavior of radium in saline formation water of the US Gulf Coast Region Isotope Geoscience 2 pp 153shy174 1984

Landreacute V Comments on ldquoRadiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Nuclear Metals Superfund Site CREW Technical Note 2006-01 January 6 2006

Lauria D C and J M O Godoy Abnormal high natural radium concentration in surface waters Journal of Environmental Radioactivity 61(2) pp 27-37 2002

Lauria D C R M R Almeida O Sracek Behavior of radium thorium and uranium in groundwater near the Buena Lagoon in the Coastal Zone of the State of Rio de Janeiro Brazil Environmental Geology 47 pp 11-19 200

MACTEC 2005 Radiological Review of Fall 2004 RI Analytical Data Nuclear Metals Superfund Site Concord Massachusetts September 26 2004

MampE 2005 Memorandum from Andrew Schkuta to Ed Conroy Subject Uranium Analysis comparison of alpha spectrometry and ICPMS results December 15 2005

PSOP Project Summary and Operations Plan Nuclear Metals Superfund Site Concord Massachusetts April 15 2005

QAPP 2004 Quality Assurance Project Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Szabo Z D E Rice C L MacLeod and T H Barringer Relation of distribution of radium nitrate and pesticides to agricultural land use and depth Kirkwood-Cohansey aquifer system New Jersey Coastal Plain 1990-1991 US Geological Survey Water-Resources Investigations Report 96-4165A 119 p 1997

15

bull

bull bull

bull bull

bull bull bull

bull bull

bull bull

bull

500

1500

Figure 1

Con

cent

ratio

n (m

gkg

)

ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

1000

0

0 5 10 15 20 25 30 35 40

Samples

Figure 2

Alp

ha S

pect

osco

py (m

gkg

)

Fall 2004 Total Uranium No Super-Depleted

1000

100

10

1

y = 10604x R2 = 09049

1 10 100 1000

ICPMS (mgkg)

Fall 2004 Super-Depleted Linear Regression

Attachment 1Connolly Jayme

From Taylor Mel issagepam ail epa gov Sent Wednesday December 21 2005 932AM To brucetdemaximiscom jhuntdemaximiscom Subject Fw Uranium Analysis Issues

Rad-versus-ICPMS RADvsiCPMS_soilse MEvsPRP_ICPMSP memodoc (158 dPDF (29 KB) OF (25 KB)

brucejohn here is mampes take on the rns vs alpha spec analysis discrepancy

Forwarded by MelissaG TaylorR1USEPAUS on 12212005 0930 AM

Conroy Ed ltEdConroyregrn-ea ecomcomgt To

MelissaG TaylorR1USEPAUSregEPA 12152005 0416 cc PM

Subject FW Uranium Analysis Issues

Melissa

Heres Andys take on the alpha spec vs ICPMS debate along with a couple of graphs showing correlation between the two methods and MampE split data vs PRP data

Please call me on the cell phone tomorrow to discuss

Ed

Edward A Conroy PE

Project Manager

METCALF amp EDDYJAECOM

7812246880 ph

6172832942 cell

1

ed conroym-e aecom com Attachment 1

-----Original Message----shyFrom Schkuta Andrew Sent Thursday December 15 2005 358 PM To Conroy Ed Subject Uranium Analysis Issues

Ed

Please find attached a memorandum that summarizes my review of why ICP-MS and alpha spec data may not necessarily agree

ltltRad-versus-ICPMS memodocgtgt

In addition I have attached 2 preliminary charts that have the following draft information

ltltRADvsiCPMS soilsedPDFgtgt

The RADvsiCPMS soilsed chart is a draft plot of MampEs alpha spec versus ICP-MS results when data was available for both samples It shows the same general trend as the PRP data with fairly good agreement between the two analyses There appears to be a slight positive bias for alpha spec data at low concentrations but this is expected as discussed in my memo

ltltMEvsPRP ICPMSPDFgtgt

The MEvsPRP ICPMS is a draft plot of MampE vs the PRPs total uranium results as measured by ICP-MS -With the exception of a few outliers which will be further investigated there appears to be good agreement between MampEs and the PRPs data As a note the median percent difference between MampEs and the PRPs data was 1

Andrew Schkuta CHMM

Technical Specialist - Chemistry

METCALF amp EDDYJAECOM

701 Edgewater Drive

Wakefield MA 01880

(781) 224-6353

(781) 245-6293 fax

2

(See attached file Rad-versus-ICPMS Attachment 1memodoc) (See attached file RADvsiCPMS soilsedPDF) (See attached file MEvsPRP ICPMSPDF)

3

Attachment 1

Metcalf amp Eddy Inc

METCALFampEDDY

Memorandum

Date December 15 2005

To Ed Conroy

From Andrew Schkuta

Subject Uranium Analysis comparison of alpha spectrometry and ICP-MS results

File WA151 Nuclear Metals Distribution

Summary This memorandum discusses the differences between uranium analysis conducted by two different methods alpha spectroscopy (alpha spec) and inductively coupled plasma shymass spectrometry (ICP-MS) In summary the alpha spec data is expected to be biased high compared to the ICP-MS data but this is not indicative that either method is being performed incorrectly

In the September 25 2005 technical memorandum Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts (MACTEC 2005) data was presented comparing alpha spec and ICP-MS analysis for uranium analysis and it was proposed that future uranium analysis be performed by ICP-MS only For samples analyzed that were collected through Spring 2005 all samples were analyzed by ICP-MS with a subset (20) analyzed by alpha spec It was noted that there appeared to be a positive bias in the alpha spec data compared to the ICP-MS data and some concern has been voiced if analysis by ICP-MS only would meet project objectives This memorandum provides background on why the two analytical methods may yield differing results and discusses how they may impact project quality objectives

There are two primary reasons why alpha spec is expected to yield results that may be positively biased

1 Sample Preparation The sample preparation method for alpha spec is more rigorous The preparation method in alpha spec breaks down the silica mineral structure of the sample By design the ICP-MS sample preparation does not break down the mineral structure and is intended to remove all metals that could become environmentally available

2 Sample Quanitation The results for alpha spec are adjusted for analyte loss by correcting the result based on the amount of tracer recovered (yield) ICP-MS data is absolute ie data is not corrected for loss during sample preparation

2 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

These two reasons are discussed in further detail as follows

Sample Preparation

Radionuclide analysis traditionally uses a sample preparation method that includes hydrofluoric acid (HF) to break down the sample so that all of the uranium present in the sample is put into solution On the other hand standard EPA methods for metals analysis including ICP-MS do not measure the total metals present in a sample For example method SW846-3050 (EPA 1996) (as specified by the PRP in their approved QAPP for metals analysis by ICP-MS) states

This method is not a total digestion technique for most samples It is a very strong acid digestion that will dissolve almost all elements that could become environmentally available By design elements bound in silicate structures are not normally dissolved by this procedure as they are not usually mobile in the environment If absolute total digestion is required use Method 3052

As noted by the method a SW846 method is available to do a total digestion (equivalent to the alpha spec method) as method 3052 However method 3052 (EPA 1996) states (emphasis added)

Note This technique is not appropriate for regulatory applications that require the use of leachate preparations (ie Method 3050 Method 3051 Method 1311 Method 1312 Method 1310 Method 1320 Method 1330 Method 3031 Method 3040) This method is appropriate for those applications requiring a total decomposition for research purposes (ie~ geological studiesr mass balancesr analysis of Standard Reference Materials) or in response to a regulation that requires total sample decomposition

Even though the ICP-MS preparation method does not yield total results it still may conservatively measure the amount of metals that are environmentally available In the Research Project Summary Ambient Levels of Metals in New Jersey Soils prepared by the New Jersey Department of Environmental Protection (NJDEP 2003) where samples were prepared using SW846-3050 procedures it states

Metal concentrations reported in this summary were acid-extractable metals not necessarily total metal concentrations The extraction method (USEPA Method 3050) is a vigorous extraction method designed to remove all metals that could become environmentally available In practice the extraction method likely overestimates metals that could become available since the sample is refluxed with both concentrated nitric acid and hydrogen peroxide However by design the method will not extract chemicals incorporated in silica minerals as they are usually not mobile in the environment Thus concentrations reported in these studies may be lower than those indicated from analyses using methods designed to measure total metal concentrations such as x-ray fluorescence methods

Although references to uranium were not located it is expected that uranium will have some of the same type of mobility characteristics as other metals in the soil matrix The paper Ground Water Issue Behavior of Metals in Soil published by EPAs Superfund Technology Support Center for Ground Water (EPA 1992) describes how metals are distributed within soil The paper describes how metals are partitioned in various pools and how a rigorous extraction using HF is required to release metals present in the structure of primary or secondary minerals It is

3 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

expected that native metals (ie background concentrations) can be present in these structures The paper indicates that for four metals examined (cadmium copper zinc and nickel) the amount of metals that was bound in the mineral structure ranges from approximately 15-60 It would be expected that naturally occurring uranium would have similar characteristics

The National Institutes of Standards amp Technology (NIST) recognizes the differences between total digestion and method SW846-3050 and has issued Addendum to Certificates for the Certificate of Analysis (NIST 2003) for some Standard Reference Materials (SRMs) The Addendum to Certificates for SRM 2711 a Montana agricultural soil indicates that trace metals were measured at approximately 40 to 90 of their certified value when analyzed using SW846-3050 as the preparation method Even though uranium is not listed as a target analyte it is reasonable to expect that is would behave in a similar fashion

Sample Quanitation

The preparation method for alpha spec analysis includes addition of a tracer (U-232) to sample and is carried through the entire analytical scheme The amount of tracer is measured during analysis and the results for target compounds (the other uranium isotopes) are corrected for the amount of tracer recovery (yield) For example if amount of tracer measured is 50 of what was initially added to the sample the sample results are all corrected by dividing by 050 In terms of method quality assurancequality control (QAQC) the PRP QAPP indicates that tracer recovery in the range of 15-125 is considered acceptable and the analytical system is considered in control In other words up to 85 of the compound can be lost during sample preparation and analysis and the data is considered meeting all QAQC objectives This method of analysis is the equivalent to the technique of isotope dilution which is used for some non-radiological environmental analyses (eg EPA Method 1625 Semivolatile Organic Compounds by Isotope Dilution GCMS) Analyses using this technique although considered the most accurate are not widely employed due to the complexity and expense of using the isotope dilution technique

ICP-MS analysis measures the concentration of analyte with no correction for potential losses However QAQC procedures are in place to make sure the losses are not considered significant In addition to laboratory control samples matrix spike analysis measures the recovery for the target analyte that was over-spiked into the sample The PRP QAPP states that matrix spike recovery should be within the range of 75-125 Therefore up to 25 of the analyte could be lost and the method has still met the QAQC objectives If recovery is less than this then validation procedures acknowledge that the data may be biased low and sample results are flagged as estimated

Therefore because the alpha spec data is corrected for analyte loss during the analytical process while ICP-MS data is not some amount of positive bias is expected to be present in alpha spec data when compared to ICP-MS data

Considerations

In evaluating which analysis is appropriate for samples from the project neither method is inherently better than the other The following factors should be considered

1 What is being measured ICP-MS measures mass directly and data can be mathematically converted to activity Alpha spec measures activity directly and can be mathematically converted to mass

4 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

2 What are the objectives of the data in terms of measuring total versus environmentally available The ICP-MS method specified in the PRP QAPP measures uranium concentrations in the same manner that EPA has measured other metals concentrations Should uranium mass be measured in a typical way that will be used for measuring other metals of concern at the site (eg copper beryllium arsenic etc)

3 If ICP-MS is more cost effective that alpha spec should the option of using preparation method SW846-3052 be considered This should yield total uranium data similar to alpha spec analysis while still using the efficiencies of ICP-MS analysis

4 What units will the cleanup level for the site be in If its in mass units then ICPshyMS analysis may be more appropriate If its in activity units then alpha spec analysis may be warranted

References

Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts Technical Memorandum from Heath Downey to Bruce Thompson de maxim is inc September 26 2005 (MACTEC 2005)

Test Methods for Evaluating Solid Waste PhysicalChemical Methods SW-846 Third Edition with updates EPA 1996

Ambient Levels of Metals in New Jersey Soils Environmental Assessment and Risk Analysis Element Research Project Summary Division of Science Research and Technology State of New Jersey Department of Environmental Protection May2003

Ground Water Issue Behavior of Metals in Soil EPA Superfund Technology Support Center for Ground Water EPA540S-92018 October 1993

Certificate of Analysis Standard Reference Material 2711 Montana Soil National Institute of Standards amp Technology July 18 2003

bull bull

bull bull bullbull bull

10000

MampE SoilSediment Results Total Uranium mgKg

bull

bull - ~-~-----~~---~~-~--~~1000

en aE

I I bull Total a u

I -

bull bull 100 1shy -middot- -middot bullbull bull

Ybull bull

10 100 1000 10000

Alpha Spec

Attachm

ent 1

10 -

100000

TOTAL URANIUM BY ICP-MS mgKg

bull

bull 10000

I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 7: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

de maximis inc

It is important to understand how the RI data will be evaluated and interpreted moving forward in the CERCLA process with respect to the concern that some areas might be missed for remediation The RI data set will be used to derive human exposure point concentrations (EPCs) for COPCs within the various media and exposure areas at the Site EPCs will be derived in accordance with EPA guidance which generally use 95 upper confidence level concentrations for soil analytes For uranium this will be performed using the ICPMS data set As a component of the risk assessment uncertainties associated with the analytical data are evaluated evaluation of alpha spectroscopy outliers as well as evaluation of the effect of uranium enrichment (eg ranging from 02 to 07) on risk estimates will be part of this process

If based on this process a given exposure areamedia has risks that EPA deems to be unacceptable (generally gt 10-4 cancer risk or non-cancer hazard index gt1 based on the cumulative risk from all COPCs in a given mediumexposure area) then preliminary remediation goals (PRGs) will be developed for the COPCs that contribute most substantially to the total risk In turn the PRGs will then be used to evaluate areas and volumes for which remedial alternatives will be considered in the Feasibility Study (FS) Areas for remediation will consider a variety of factors when determining the overall dimensions and they typically will be drawn to include RI data points that are equal to or greater than the PRGs

EPA will then select final remediation goals in the Record of Decision (ROD) The Remedial DesignRemedial Action (RDRA) process occurs after the ROD During the RD areas and volumes for remediation will be revisited based on the final remediation goals typically resulting in identification of a minimum target areavolume that will be remediated As part of the RD process a Demonstration of Compliance Plan will be developed This document sets forth how the contractor will demonstrate that all targeted media has been successfully remediated For a site like NMI we would expect that different instrumentation analytical methods may be used on Site to evaluate the effectiveness of remediation during the RA (eg gamma spectroscopy or XRF which provides faster results than ICPMS or alpha spectroscopy) followed by confirmatory sampling sent to an off-site laboratory Correlation of these instrument results to laboratory analytical methods will be presented at the appropriate time Given this process (CERCLA) and the multiple reviews by EPA State of Massachusetts agencies and public groups it is very unlikely that any areas with unacceptable risk levels will be missed for remediation

With respect to uranium background concentrations the total uranium values calculated using alpha spectroscopy values is correct in the CREW Technical Note (2006) however this background value is based on a theoretical calculation While this calculation is correct de maximis inc (and EPA via MampE split samples) have shown total uranium (reflected in the background evaluation) directly measured by ICPMS is lower due to inherent differences between the two methods and the fact that ICPMS is measuring ldquoenvironmentally availablerdquo concentrations Furthermore the background concentration at the Site is not likely to play a significant role since risk-based remediation levels greater than background are expected to be established for Site clean-up

7

de maximis inc

Lastly the intent for referencing the urine study was to show an independent evaluation that also concludes ICPMS has lower sensitivity for uranium and a greater ability to distinguish between natural depleted and enriched uranium Further it supports the argument that some variability between ICPMS and alpha spectroscopy analysis is not unexpected

In conclusion the ICPMS analysis provides lower results as compared to alpha spectroscopy analysis results however as discussed in this response (and described in detail in MampE uranium evaluation memo) neither method is in error

bull Background uranium concentrations determined from actual ICPMS measurements show environmentally available U background is approximately 13 ppm (EPA MampE-split samples confirm this)

bull lsquoSuper-depletedrsquo points are anomalies and these samples will be recollected and reanalyzed however remaining total U concentrations are usable and ICPMS results generated for the RIFS are reliable

bull Urine analysis referenced in the Technical Memorandum is applicable in that it supports the use of ICPMS for evaluating presence of U235 and calculating enrichments (for future samples once alpha spectroscopy analysis is discontinued)

bull Alpha spectroscopy analysis analyses for the Site were intended to verify enrichment to ensure that no enriched uranium was missed by ICPMS and to confirm signature of uranium and thorium at the site This has been done therefore future alpha spectroscopy analysis is not needed

Response to Section 3 ndash Using enrichment rather than total U for determination of contamination

Using total uranium by mass rather than enrichment to define nature and extent of contamination is being performed at this Site because the determination of enrichment at total uranium concentrations in the background range (eg 075 ndash 25 mgkg) is uncertain particularly for alpha spectroscopy analysis methods where the total propagated uncertainty of measurements is nearly equal to the actual instrument reading This is clearly demonstrated in the scatter observed in data plots provided by CREW (2006) for total uranium in this range de maximis inc concurs that at 5 mgkg a large percentage of the mass should be depleted uranium and for the purposes of defining the nature and extent of contamination based solely on isotopic composition a value of 5 would be inappropriate However the RI is using a value of 21 mgkg as a threshold for the definition of the nature and extent of contamination The nature and extent of uranium like all substances at the Site is being defined by risk-based levels (background levels are only used when background values are higher than risk-based levels) Therefore for total uranium concentrations less than the RI screening level the

8

de maximis inc

source of the uranium (whether depleted or natural uranium) is not relevant For risk assessment purposes all uranium will be considered and the contribution to risk associated with background constituents of concern will be considered after the total risk is calculated

Measurements of environmental data rarely have a normal (Gaussian) distribution and typically are lognormal distributions (Gilbert 1987) EPA guidance indicates that ldquoit is unusual to encounter environmental data sets that are normally distributedrdquo (EPA 2002) and ldquoenvironmental data commonly exhibit frequency distributions that are non-negative and skewed with heavy or long right tailsrdquo (EPA 2000) The composition of uranium (natural or depleted) will be used during risk assessment but it has limited usefulness otherwise However the isotopic composition of uranium will be one of the issues considered during data evaluation to avoid the sort of issues pointed out in the CREW Technical Note

Response to Section 4 - Thorium evaluation

de maximis inc re-iterates that the Technical Memorandum shows thorium-232 and thorium-228 concentrations are roughly equal (and considered within the analytical variability for alpha spectroscopy) and that thorium-230 at the Site is not related to Site activities Furthermore uranium-232 (from recycled uranium) is not present at the site as it is primarily a concern with enriched uranium Even if present it would not be in significant enough quantities to impact thorium-228 concentrations

Refer to previous discussion pertaining to ICPMS versus alpha spectroscopy analysis information Also note that ICPMS instrument sensitivity for thorium is very high due to the high mass to charge ratio (ie it is a good responder) Naturally occurring thorium is 100 Th-232 No data quality issues were observed during data validation and data has been deemed usable

Lastly thorium was used at the Site but at significantly lower amounts when compared to uranium Therefore low concentrations (within background concentration range) of thorium are not unexpected and historical data have demonstrated this

Response to Section 5 - Other radionuclides

There is no way to know the exact pedigree of the depleted uranium processed at the Site or to determine which portion of the depleted uranium might have come from reprocessed fuel rods and therefore contain small amounts of reactor generated radionuclides Studies of depleted uranium have shown that byproduct radionuclides in uranium are present at very low levels In addition dose assessments of these trace contaminants have shown that the radiological dose from these radionuclides is insignificant compared to that from uranium These conclusions were documented in a January 2000 Department of the Army analysis of transuranics and other contaminants in depleted uranium armor (Attachment 3) Of particular relevance is the fact that the Department of the Army analysis took samples of DU billets from Starmet for this study The Army concluded that ldquoThe levels of transuranics elements (TRUs) and fission

9

de maximis inc

products (FP) are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is saferdquo If the TRUs and FP contribute less than a one percent increase in dose for DU metal then they would contribute an even more negligible dose at the 10s to 1000s of mgkg of DU present in NMI Site media With respect to the HTDR analysis during the RI historical data indicated that the holding basin and former drain outfall regions of the bog were the most likely areas to have significant concentrations of uranium therefore the RI work plan identified locations to be sampled in these areas (FSP 2004) Further the approach outlined in the RI was to evaluate for HTDR in samples with elevated concentrations of uranium not necessarily the highest concentrations There were other AOIs where uranium was detected in elevated concentrations but no sample was collected for HTDR However since the holding basin was in use for the vast majority of time that uranium was actively processed at the site and no HTDR were detected in the six samples from this area it is unlikely that HTDR would be identified in other areas of the site

Samples for HTDRs were collected according to the RI (FSP 2004) such that 3 of the samples from AOI 1 were above the water table and 3 of the samples were from below the water table one sample was collected from each of the AOI 6 former drain outfall areas of the bog and one sample collected from the buried drum excavation (AOI 2) These HTDR samples are representative for the areas investigated (AOI 1 AOI 2 and AOI 6) and of the site based on the distribution of total uranium results described below

bull Total uranium in the HTDR sample locations from AOI 1 ranged from 38 to 511 mgkg while total uranium for all samples in AOI 1 ranged from 11 to 672 mgkg Only 25 samples from all samples in AOI 1 had total uranium concentrations greater than 100 mgkg and 4 of the 6 HTDR samples were from this subset

bull Total uranium in the HTDR sample locations from AOI 6 were 25 mgkg at former drain outfall 1 and 4550 mgkg at drain outfall 2 while total uranium for all samples in AOI 6 ranged from 16 to 4550 mgkg Only 10 samples from all samples in AOI 6 had total uranium concentrations greater than 100 mgkg and one HTDR sample was from the maximum total uranium sample location

bull The HTDR sample from AOI 2 drum excavation bottom had a total uranium concentration of 18 mgkg while total uranium for all excavation sides and bottom ranged from 14 to 49 mgkg

Only a few samples have uranium concentrations greater than 1000 mgkg and 6 of the 8 are from samples deemed as lsquosuper-depletedrsquo These 6 locations will be re-sampled as part of Phase 1C since the ICPMS analysis results are considered an anomaly Sporadic low detections of HTDRs do not necessarily indicate the presence of HTDRs as false positive results are common for these analyses Although some HTDRs have been detected in the past they are sporadic (not all samples have detections) and do not indicate the presence of HTDRs at significant levels

10

de maximis inc

It is also important to note that the historical data provided in the RI (PSOP 2005) was used to assist in design of the RI (FSP 2004) Data were compiled from many sources of information and represent samples collected analyzed and reported using various sample collection procedures laboratory analytical methods and data reporting conventions Data compilation activities involved both manual hand-entries and electronic entries when applicable Based on a review of the electronic database severe data quality limitations were noted including unknown analytical quality and validation levels missing location coordinates inconsistent media codes unknown detection limits and radionuclide uncertainty measurements inconsistent andor incorrect reporting of radionuclides using mass-based and activity-based units These limitations were considered significant during the scoping of the RIFS Work Plan therefore the historical data were used primarily to evaluate the spatial distribution of parameters and magnitude of detected concentrations The limitations of the historical data were used to provide the basis for identifying data gaps and scoping the RIFS As such the FSP included analysis for HTDRs at the areas of the Site known to be impacted by historical practices

With respect to comments provided on U-232 and U-236 these uranium isotopes are produced during the fission of uranium (as are plutonium americium and the rest of the HTDRs) Since none of these other fission products have been detected at the Site it is not credible that either uranium-232 or uranium-236 would be present

Response to Section 6 - Radium

The historical radium in groundwater data is suspect for several reasons including the previous discussion on historical data provided in the RI (PSOP 2005) Groundwater samples were collected from 13 wells at the Site (single sampling event) using bailers as part of this investigation All samples were screened for radioactivity by gross alpha beta analysis Only one sample exceeded the action level (gt 10 pCil gross alpha) with results of 50 pCil gross alpha and 55 pCil gross beta and it was analyzed for isotopic uranium Ra-226 and Ra-228 The results were 74 pCil U-234 025 pCil U-235 037 pCil U-238 12 pCil Ra-226 and 7 pCil Ra-228 These values are from Table 10 of the report and it appears that the Ra-226 value was incorrectly transcribed in the CREW Technical Note as 37 pCil The isotopic uranium results do not match those expected as U-234 is elevated and results for U-235 and U-238 are almost equal Therefore the analytical results for this sample are questionable Furthermore all the rest of the ground water samples had gross alpha results of 3 pCil or less indicating that there are no significant amounts of radium in groundwater

In addition a limited literature review was performed to assess the potential for leaching of radium from subsurface solids by nitrate at the NMI Site The literature demonstrates that radium may be found in a handful of aqueous phase complexes The predominant species of radium in aqueous solution is the divalent cation Ra2+ but other complexes such as RaOH+ RaCl+ RaCO3 and RaSO4 may also be found (Beneš et al 1982 Beneš 1984) Radium appears to be the least likely of all the alkali earth metals (beryllium magnesium calcium strontium barium and radium) to form aqueous phase

11

de maximis inc

complexes (Kirby and Salutsky 1964) As such we are unclear as to the origin of Dr Landreacutersquos statement that ldquonitrates are known to influence greatly radium leaching from the soil into the groundwaterrdquo (Landreacute 2006) The salt Ra(NO3)2 is highly soluble (245 g100 g H2O Kirby and Salutsky 1964) but we have not found evidence to suggest that the species Ra(NO3)+ or Ra(NO3)2 exist in aqueous solution Lauria et al (2004) present radionuclide and water chemistry data from groundwater samples taken from the Buena Lagoon in Brazil A plot of Ra-226 and Ra-228 concentrations versus nitrate concentration is presented in Figure 1 These data demonstrate that both Ra-226 and Ra-228 are almost completely uncorrelated with nitrate at this site as evidenced by correlation coefficients (r2) of approximately 005 and 004 respectively Szabo et al (1997) also report that no correlation exists between nitrate and Ra-226 in 17 groundwater samples taken from the Kirkwood-Cohansey aquifer system in New Jersey although Ra-228 appears to be weakly correlated with nitrate in the same data set (r2 = 018)

Figure 1 Correlation between aqueous radium and nitrate (Lauria et al 2004)

r 2 = 005

r 2 = 004

1E-10 1E-09 1E-08 1E-07 1E-06 1E-05 1E-04

0 5 10 15 20 25

Nitrate (mgL)

Ra

(microg

L)

Ra-226 Ra-228 Linear (Ra-226) Linear (Ra-228)

The mobility of radium in the subsurface appears to be controlled primarily by sorption and precipitation (Lauria et al 2004) High values of aqueous phase radium have been observed in groundwater samples with either increased salinity or reduced pH (Kraemer and Reed 1984 Lauria and Godoy 2002) Demir (1989) demonstrated the increased mobility of radium in acidic environments due to the reduced sorptive capacity of sediments at low pH Figure 2 is a plot of radium concentration versus sample pH from the same data set (Lauria et al 2004)

12

de maximis inc

Figure 2 Correlation between aqueous radium and pH (Lauria et al 2004) 226Ra = -4E-06pH + 3E-05

r2 = 058

228Ra = -5E-08pH + 4E-07 r2 = 032

1E-10 1E-09 1E-08 1E-07 1E-06 1E-05 1E-04

3 4 5 6 7

pH

Ra

(microg

L)

Ra-226 Ra-228 Linear (Ra-226) Linear (Ra-228)

Here a stronger correlation between Ra-226 concentration and pH exists (r2 = 058) along with an observable trend of decreasing aqueous phase radium with increasing pH Szabo et al (1997) also report a negative correlation (r2 = 030) between Ra-228 and pH but no significant correlation between Ra-226 and pH Divalent cations such as calcium and magnesium compete with radium for a limited number of surface complexation and adsorption sites (Szabo et al 1997) As lime (a mixture potentially containing calcium oxide calcium hydroxide and calcium carbonate) was co-disposed with uranium and nitric acid at the Holding Basin nitrate may serve as a site-specific tracer for conditions that may lead to the release of adsorbed or surface-complexed radium from the solid phase although no comparable calcium ldquoplumerdquo has been defined unlike nitrate in groundwater Therefore we suggest that Dr Landreacutersquos statement could be qualified by stating that low pH values are known to influence the leaching of radium from aquifer solids However as discussed above we do not see a compelling reason to add radium to the analytical list at this time

Please contact me if you have any questions

Sincerely

Bruce Thompson

cc Respondents Bennie Underwood PE de maximis inc inc Nelson Walter PE MACTEC Inc Peter Zeeb PhD PG GeoSyntec Consultants Inc

13

de maximis inc

Figure 1 ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

Figure 2 Comparison of Alpha Spectroscopy and ICP-MS Uranium Analysis

Attachment 1 Metcalf amp Eddy Memorandum ndash Uranium Analysis comparison of alpha spectrometry and ICP-MS results

Attachment 2 QAPP SOPs for Alpha Spectroscopy and ICPMS Sample Preparation

Attachment 3 Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor Department of the Army January 19 2000

References

Beneš P Migration of radium in the terrestrial hydrosphere The behavior of radium in waterways and aquifers Technical Reports Series No 310 International Atomic Energy Agency IAEA-TECDOC-301 Vienna Austrai pp 119-173 1984

Beneš P M Obdržaacutelek M Čejchanovaacute The physicochemical forms of tracers of radium in aqueous solutions containing chlorides sulfates and carbonates Radiochem Radioanal Letters 50(4) pp 227-242 1982

CREW 2006 ldquoComments on Radiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Technical Note No 2006-01 January 6 2006

Demir I Temporal variations and sources of Pb Co Cu Ni Fe and Mo in shallow ground water of the McDonalds Branch basin Lebanon State Forest Burlington County New Jersey New Jersey Geological Survey Report 20 19 p 1989

EPA 2000 Guidance for Data Quality Assessment Practical Methods for Data Analysis EPA QAG-9 July 2000

EPA 2002 Guidance for Comparing Background and Chemical Concentrations in Soil for CERCLA Site EPA 540-R-01-003 September 2002

FSP 2004 Field Sampling Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Gilbert 1989 Statistical Methods for Environmental Pollution Monitoring Richard O Gilbert 1987

Kirby H W M L Salutsky The Radiochemistry of Radium National Academy of Sciences ndash National Research Council NAS-NS 3057 205 p 1964

14

de maximis inc

Kramer T F and D F Reid The occurrence and behavior of radium in saline formation water of the US Gulf Coast Region Isotope Geoscience 2 pp 153shy174 1984

Landreacute V Comments on ldquoRadiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Nuclear Metals Superfund Site CREW Technical Note 2006-01 January 6 2006

Lauria D C and J M O Godoy Abnormal high natural radium concentration in surface waters Journal of Environmental Radioactivity 61(2) pp 27-37 2002

Lauria D C R M R Almeida O Sracek Behavior of radium thorium and uranium in groundwater near the Buena Lagoon in the Coastal Zone of the State of Rio de Janeiro Brazil Environmental Geology 47 pp 11-19 200

MACTEC 2005 Radiological Review of Fall 2004 RI Analytical Data Nuclear Metals Superfund Site Concord Massachusetts September 26 2004

MampE 2005 Memorandum from Andrew Schkuta to Ed Conroy Subject Uranium Analysis comparison of alpha spectrometry and ICPMS results December 15 2005

PSOP Project Summary and Operations Plan Nuclear Metals Superfund Site Concord Massachusetts April 15 2005

QAPP 2004 Quality Assurance Project Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Szabo Z D E Rice C L MacLeod and T H Barringer Relation of distribution of radium nitrate and pesticides to agricultural land use and depth Kirkwood-Cohansey aquifer system New Jersey Coastal Plain 1990-1991 US Geological Survey Water-Resources Investigations Report 96-4165A 119 p 1997

15

bull

bull bull

bull bull

bull bull bull

bull bull

bull bull

bull

500

1500

Figure 1

Con

cent

ratio

n (m

gkg

)

ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

1000

0

0 5 10 15 20 25 30 35 40

Samples

Figure 2

Alp

ha S

pect

osco

py (m

gkg

)

Fall 2004 Total Uranium No Super-Depleted

1000

100

10

1

y = 10604x R2 = 09049

1 10 100 1000

ICPMS (mgkg)

Fall 2004 Super-Depleted Linear Regression

Attachment 1Connolly Jayme

From Taylor Mel issagepam ail epa gov Sent Wednesday December 21 2005 932AM To brucetdemaximiscom jhuntdemaximiscom Subject Fw Uranium Analysis Issues

Rad-versus-ICPMS RADvsiCPMS_soilse MEvsPRP_ICPMSP memodoc (158 dPDF (29 KB) OF (25 KB)

brucejohn here is mampes take on the rns vs alpha spec analysis discrepancy

Forwarded by MelissaG TaylorR1USEPAUS on 12212005 0930 AM

Conroy Ed ltEdConroyregrn-ea ecomcomgt To

MelissaG TaylorR1USEPAUSregEPA 12152005 0416 cc PM

Subject FW Uranium Analysis Issues

Melissa

Heres Andys take on the alpha spec vs ICPMS debate along with a couple of graphs showing correlation between the two methods and MampE split data vs PRP data

Please call me on the cell phone tomorrow to discuss

Ed

Edward A Conroy PE

Project Manager

METCALF amp EDDYJAECOM

7812246880 ph

6172832942 cell

1

ed conroym-e aecom com Attachment 1

-----Original Message----shyFrom Schkuta Andrew Sent Thursday December 15 2005 358 PM To Conroy Ed Subject Uranium Analysis Issues

Ed

Please find attached a memorandum that summarizes my review of why ICP-MS and alpha spec data may not necessarily agree

ltltRad-versus-ICPMS memodocgtgt

In addition I have attached 2 preliminary charts that have the following draft information

ltltRADvsiCPMS soilsedPDFgtgt

The RADvsiCPMS soilsed chart is a draft plot of MampEs alpha spec versus ICP-MS results when data was available for both samples It shows the same general trend as the PRP data with fairly good agreement between the two analyses There appears to be a slight positive bias for alpha spec data at low concentrations but this is expected as discussed in my memo

ltltMEvsPRP ICPMSPDFgtgt

The MEvsPRP ICPMS is a draft plot of MampE vs the PRPs total uranium results as measured by ICP-MS -With the exception of a few outliers which will be further investigated there appears to be good agreement between MampEs and the PRPs data As a note the median percent difference between MampEs and the PRPs data was 1

Andrew Schkuta CHMM

Technical Specialist - Chemistry

METCALF amp EDDYJAECOM

701 Edgewater Drive

Wakefield MA 01880

(781) 224-6353

(781) 245-6293 fax

2

(See attached file Rad-versus-ICPMS Attachment 1memodoc) (See attached file RADvsiCPMS soilsedPDF) (See attached file MEvsPRP ICPMSPDF)

3

Attachment 1

Metcalf amp Eddy Inc

METCALFampEDDY

Memorandum

Date December 15 2005

To Ed Conroy

From Andrew Schkuta

Subject Uranium Analysis comparison of alpha spectrometry and ICP-MS results

File WA151 Nuclear Metals Distribution

Summary This memorandum discusses the differences between uranium analysis conducted by two different methods alpha spectroscopy (alpha spec) and inductively coupled plasma shymass spectrometry (ICP-MS) In summary the alpha spec data is expected to be biased high compared to the ICP-MS data but this is not indicative that either method is being performed incorrectly

In the September 25 2005 technical memorandum Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts (MACTEC 2005) data was presented comparing alpha spec and ICP-MS analysis for uranium analysis and it was proposed that future uranium analysis be performed by ICP-MS only For samples analyzed that were collected through Spring 2005 all samples were analyzed by ICP-MS with a subset (20) analyzed by alpha spec It was noted that there appeared to be a positive bias in the alpha spec data compared to the ICP-MS data and some concern has been voiced if analysis by ICP-MS only would meet project objectives This memorandum provides background on why the two analytical methods may yield differing results and discusses how they may impact project quality objectives

There are two primary reasons why alpha spec is expected to yield results that may be positively biased

1 Sample Preparation The sample preparation method for alpha spec is more rigorous The preparation method in alpha spec breaks down the silica mineral structure of the sample By design the ICP-MS sample preparation does not break down the mineral structure and is intended to remove all metals that could become environmentally available

2 Sample Quanitation The results for alpha spec are adjusted for analyte loss by correcting the result based on the amount of tracer recovered (yield) ICP-MS data is absolute ie data is not corrected for loss during sample preparation

2 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

These two reasons are discussed in further detail as follows

Sample Preparation

Radionuclide analysis traditionally uses a sample preparation method that includes hydrofluoric acid (HF) to break down the sample so that all of the uranium present in the sample is put into solution On the other hand standard EPA methods for metals analysis including ICP-MS do not measure the total metals present in a sample For example method SW846-3050 (EPA 1996) (as specified by the PRP in their approved QAPP for metals analysis by ICP-MS) states

This method is not a total digestion technique for most samples It is a very strong acid digestion that will dissolve almost all elements that could become environmentally available By design elements bound in silicate structures are not normally dissolved by this procedure as they are not usually mobile in the environment If absolute total digestion is required use Method 3052

As noted by the method a SW846 method is available to do a total digestion (equivalent to the alpha spec method) as method 3052 However method 3052 (EPA 1996) states (emphasis added)

Note This technique is not appropriate for regulatory applications that require the use of leachate preparations (ie Method 3050 Method 3051 Method 1311 Method 1312 Method 1310 Method 1320 Method 1330 Method 3031 Method 3040) This method is appropriate for those applications requiring a total decomposition for research purposes (ie~ geological studiesr mass balancesr analysis of Standard Reference Materials) or in response to a regulation that requires total sample decomposition

Even though the ICP-MS preparation method does not yield total results it still may conservatively measure the amount of metals that are environmentally available In the Research Project Summary Ambient Levels of Metals in New Jersey Soils prepared by the New Jersey Department of Environmental Protection (NJDEP 2003) where samples were prepared using SW846-3050 procedures it states

Metal concentrations reported in this summary were acid-extractable metals not necessarily total metal concentrations The extraction method (USEPA Method 3050) is a vigorous extraction method designed to remove all metals that could become environmentally available In practice the extraction method likely overestimates metals that could become available since the sample is refluxed with both concentrated nitric acid and hydrogen peroxide However by design the method will not extract chemicals incorporated in silica minerals as they are usually not mobile in the environment Thus concentrations reported in these studies may be lower than those indicated from analyses using methods designed to measure total metal concentrations such as x-ray fluorescence methods

Although references to uranium were not located it is expected that uranium will have some of the same type of mobility characteristics as other metals in the soil matrix The paper Ground Water Issue Behavior of Metals in Soil published by EPAs Superfund Technology Support Center for Ground Water (EPA 1992) describes how metals are distributed within soil The paper describes how metals are partitioned in various pools and how a rigorous extraction using HF is required to release metals present in the structure of primary or secondary minerals It is

3 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

expected that native metals (ie background concentrations) can be present in these structures The paper indicates that for four metals examined (cadmium copper zinc and nickel) the amount of metals that was bound in the mineral structure ranges from approximately 15-60 It would be expected that naturally occurring uranium would have similar characteristics

The National Institutes of Standards amp Technology (NIST) recognizes the differences between total digestion and method SW846-3050 and has issued Addendum to Certificates for the Certificate of Analysis (NIST 2003) for some Standard Reference Materials (SRMs) The Addendum to Certificates for SRM 2711 a Montana agricultural soil indicates that trace metals were measured at approximately 40 to 90 of their certified value when analyzed using SW846-3050 as the preparation method Even though uranium is not listed as a target analyte it is reasonable to expect that is would behave in a similar fashion

Sample Quanitation

The preparation method for alpha spec analysis includes addition of a tracer (U-232) to sample and is carried through the entire analytical scheme The amount of tracer is measured during analysis and the results for target compounds (the other uranium isotopes) are corrected for the amount of tracer recovery (yield) For example if amount of tracer measured is 50 of what was initially added to the sample the sample results are all corrected by dividing by 050 In terms of method quality assurancequality control (QAQC) the PRP QAPP indicates that tracer recovery in the range of 15-125 is considered acceptable and the analytical system is considered in control In other words up to 85 of the compound can be lost during sample preparation and analysis and the data is considered meeting all QAQC objectives This method of analysis is the equivalent to the technique of isotope dilution which is used for some non-radiological environmental analyses (eg EPA Method 1625 Semivolatile Organic Compounds by Isotope Dilution GCMS) Analyses using this technique although considered the most accurate are not widely employed due to the complexity and expense of using the isotope dilution technique

ICP-MS analysis measures the concentration of analyte with no correction for potential losses However QAQC procedures are in place to make sure the losses are not considered significant In addition to laboratory control samples matrix spike analysis measures the recovery for the target analyte that was over-spiked into the sample The PRP QAPP states that matrix spike recovery should be within the range of 75-125 Therefore up to 25 of the analyte could be lost and the method has still met the QAQC objectives If recovery is less than this then validation procedures acknowledge that the data may be biased low and sample results are flagged as estimated

Therefore because the alpha spec data is corrected for analyte loss during the analytical process while ICP-MS data is not some amount of positive bias is expected to be present in alpha spec data when compared to ICP-MS data

Considerations

In evaluating which analysis is appropriate for samples from the project neither method is inherently better than the other The following factors should be considered

1 What is being measured ICP-MS measures mass directly and data can be mathematically converted to activity Alpha spec measures activity directly and can be mathematically converted to mass

4 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

2 What are the objectives of the data in terms of measuring total versus environmentally available The ICP-MS method specified in the PRP QAPP measures uranium concentrations in the same manner that EPA has measured other metals concentrations Should uranium mass be measured in a typical way that will be used for measuring other metals of concern at the site (eg copper beryllium arsenic etc)

3 If ICP-MS is more cost effective that alpha spec should the option of using preparation method SW846-3052 be considered This should yield total uranium data similar to alpha spec analysis while still using the efficiencies of ICP-MS analysis

4 What units will the cleanup level for the site be in If its in mass units then ICPshyMS analysis may be more appropriate If its in activity units then alpha spec analysis may be warranted

References

Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts Technical Memorandum from Heath Downey to Bruce Thompson de maxim is inc September 26 2005 (MACTEC 2005)

Test Methods for Evaluating Solid Waste PhysicalChemical Methods SW-846 Third Edition with updates EPA 1996

Ambient Levels of Metals in New Jersey Soils Environmental Assessment and Risk Analysis Element Research Project Summary Division of Science Research and Technology State of New Jersey Department of Environmental Protection May2003

Ground Water Issue Behavior of Metals in Soil EPA Superfund Technology Support Center for Ground Water EPA540S-92018 October 1993

Certificate of Analysis Standard Reference Material 2711 Montana Soil National Institute of Standards amp Technology July 18 2003

bull bull

bull bull bullbull bull

10000

MampE SoilSediment Results Total Uranium mgKg

bull

bull - ~-~-----~~---~~-~--~~1000

en aE

I I bull Total a u

I -

bull bull 100 1shy -middot- -middot bullbull bull

Ybull bull

10 100 1000 10000

Alpha Spec

Attachm

ent 1

10 -

100000

TOTAL URANIUM BY ICP-MS mgKg

bull

bull 10000

I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 8: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

de maximis inc

Lastly the intent for referencing the urine study was to show an independent evaluation that also concludes ICPMS has lower sensitivity for uranium and a greater ability to distinguish between natural depleted and enriched uranium Further it supports the argument that some variability between ICPMS and alpha spectroscopy analysis is not unexpected

In conclusion the ICPMS analysis provides lower results as compared to alpha spectroscopy analysis results however as discussed in this response (and described in detail in MampE uranium evaluation memo) neither method is in error

bull Background uranium concentrations determined from actual ICPMS measurements show environmentally available U background is approximately 13 ppm (EPA MampE-split samples confirm this)

bull lsquoSuper-depletedrsquo points are anomalies and these samples will be recollected and reanalyzed however remaining total U concentrations are usable and ICPMS results generated for the RIFS are reliable

bull Urine analysis referenced in the Technical Memorandum is applicable in that it supports the use of ICPMS for evaluating presence of U235 and calculating enrichments (for future samples once alpha spectroscopy analysis is discontinued)

bull Alpha spectroscopy analysis analyses for the Site were intended to verify enrichment to ensure that no enriched uranium was missed by ICPMS and to confirm signature of uranium and thorium at the site This has been done therefore future alpha spectroscopy analysis is not needed

Response to Section 3 ndash Using enrichment rather than total U for determination of contamination

Using total uranium by mass rather than enrichment to define nature and extent of contamination is being performed at this Site because the determination of enrichment at total uranium concentrations in the background range (eg 075 ndash 25 mgkg) is uncertain particularly for alpha spectroscopy analysis methods where the total propagated uncertainty of measurements is nearly equal to the actual instrument reading This is clearly demonstrated in the scatter observed in data plots provided by CREW (2006) for total uranium in this range de maximis inc concurs that at 5 mgkg a large percentage of the mass should be depleted uranium and for the purposes of defining the nature and extent of contamination based solely on isotopic composition a value of 5 would be inappropriate However the RI is using a value of 21 mgkg as a threshold for the definition of the nature and extent of contamination The nature and extent of uranium like all substances at the Site is being defined by risk-based levels (background levels are only used when background values are higher than risk-based levels) Therefore for total uranium concentrations less than the RI screening level the

8

de maximis inc

source of the uranium (whether depleted or natural uranium) is not relevant For risk assessment purposes all uranium will be considered and the contribution to risk associated with background constituents of concern will be considered after the total risk is calculated

Measurements of environmental data rarely have a normal (Gaussian) distribution and typically are lognormal distributions (Gilbert 1987) EPA guidance indicates that ldquoit is unusual to encounter environmental data sets that are normally distributedrdquo (EPA 2002) and ldquoenvironmental data commonly exhibit frequency distributions that are non-negative and skewed with heavy or long right tailsrdquo (EPA 2000) The composition of uranium (natural or depleted) will be used during risk assessment but it has limited usefulness otherwise However the isotopic composition of uranium will be one of the issues considered during data evaluation to avoid the sort of issues pointed out in the CREW Technical Note

Response to Section 4 - Thorium evaluation

de maximis inc re-iterates that the Technical Memorandum shows thorium-232 and thorium-228 concentrations are roughly equal (and considered within the analytical variability for alpha spectroscopy) and that thorium-230 at the Site is not related to Site activities Furthermore uranium-232 (from recycled uranium) is not present at the site as it is primarily a concern with enriched uranium Even if present it would not be in significant enough quantities to impact thorium-228 concentrations

Refer to previous discussion pertaining to ICPMS versus alpha spectroscopy analysis information Also note that ICPMS instrument sensitivity for thorium is very high due to the high mass to charge ratio (ie it is a good responder) Naturally occurring thorium is 100 Th-232 No data quality issues were observed during data validation and data has been deemed usable

Lastly thorium was used at the Site but at significantly lower amounts when compared to uranium Therefore low concentrations (within background concentration range) of thorium are not unexpected and historical data have demonstrated this

Response to Section 5 - Other radionuclides

There is no way to know the exact pedigree of the depleted uranium processed at the Site or to determine which portion of the depleted uranium might have come from reprocessed fuel rods and therefore contain small amounts of reactor generated radionuclides Studies of depleted uranium have shown that byproduct radionuclides in uranium are present at very low levels In addition dose assessments of these trace contaminants have shown that the radiological dose from these radionuclides is insignificant compared to that from uranium These conclusions were documented in a January 2000 Department of the Army analysis of transuranics and other contaminants in depleted uranium armor (Attachment 3) Of particular relevance is the fact that the Department of the Army analysis took samples of DU billets from Starmet for this study The Army concluded that ldquoThe levels of transuranics elements (TRUs) and fission

9

de maximis inc

products (FP) are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is saferdquo If the TRUs and FP contribute less than a one percent increase in dose for DU metal then they would contribute an even more negligible dose at the 10s to 1000s of mgkg of DU present in NMI Site media With respect to the HTDR analysis during the RI historical data indicated that the holding basin and former drain outfall regions of the bog were the most likely areas to have significant concentrations of uranium therefore the RI work plan identified locations to be sampled in these areas (FSP 2004) Further the approach outlined in the RI was to evaluate for HTDR in samples with elevated concentrations of uranium not necessarily the highest concentrations There were other AOIs where uranium was detected in elevated concentrations but no sample was collected for HTDR However since the holding basin was in use for the vast majority of time that uranium was actively processed at the site and no HTDR were detected in the six samples from this area it is unlikely that HTDR would be identified in other areas of the site

Samples for HTDRs were collected according to the RI (FSP 2004) such that 3 of the samples from AOI 1 were above the water table and 3 of the samples were from below the water table one sample was collected from each of the AOI 6 former drain outfall areas of the bog and one sample collected from the buried drum excavation (AOI 2) These HTDR samples are representative for the areas investigated (AOI 1 AOI 2 and AOI 6) and of the site based on the distribution of total uranium results described below

bull Total uranium in the HTDR sample locations from AOI 1 ranged from 38 to 511 mgkg while total uranium for all samples in AOI 1 ranged from 11 to 672 mgkg Only 25 samples from all samples in AOI 1 had total uranium concentrations greater than 100 mgkg and 4 of the 6 HTDR samples were from this subset

bull Total uranium in the HTDR sample locations from AOI 6 were 25 mgkg at former drain outfall 1 and 4550 mgkg at drain outfall 2 while total uranium for all samples in AOI 6 ranged from 16 to 4550 mgkg Only 10 samples from all samples in AOI 6 had total uranium concentrations greater than 100 mgkg and one HTDR sample was from the maximum total uranium sample location

bull The HTDR sample from AOI 2 drum excavation bottom had a total uranium concentration of 18 mgkg while total uranium for all excavation sides and bottom ranged from 14 to 49 mgkg

Only a few samples have uranium concentrations greater than 1000 mgkg and 6 of the 8 are from samples deemed as lsquosuper-depletedrsquo These 6 locations will be re-sampled as part of Phase 1C since the ICPMS analysis results are considered an anomaly Sporadic low detections of HTDRs do not necessarily indicate the presence of HTDRs as false positive results are common for these analyses Although some HTDRs have been detected in the past they are sporadic (not all samples have detections) and do not indicate the presence of HTDRs at significant levels

10

de maximis inc

It is also important to note that the historical data provided in the RI (PSOP 2005) was used to assist in design of the RI (FSP 2004) Data were compiled from many sources of information and represent samples collected analyzed and reported using various sample collection procedures laboratory analytical methods and data reporting conventions Data compilation activities involved both manual hand-entries and electronic entries when applicable Based on a review of the electronic database severe data quality limitations were noted including unknown analytical quality and validation levels missing location coordinates inconsistent media codes unknown detection limits and radionuclide uncertainty measurements inconsistent andor incorrect reporting of radionuclides using mass-based and activity-based units These limitations were considered significant during the scoping of the RIFS Work Plan therefore the historical data were used primarily to evaluate the spatial distribution of parameters and magnitude of detected concentrations The limitations of the historical data were used to provide the basis for identifying data gaps and scoping the RIFS As such the FSP included analysis for HTDRs at the areas of the Site known to be impacted by historical practices

With respect to comments provided on U-232 and U-236 these uranium isotopes are produced during the fission of uranium (as are plutonium americium and the rest of the HTDRs) Since none of these other fission products have been detected at the Site it is not credible that either uranium-232 or uranium-236 would be present

Response to Section 6 - Radium

The historical radium in groundwater data is suspect for several reasons including the previous discussion on historical data provided in the RI (PSOP 2005) Groundwater samples were collected from 13 wells at the Site (single sampling event) using bailers as part of this investigation All samples were screened for radioactivity by gross alpha beta analysis Only one sample exceeded the action level (gt 10 pCil gross alpha) with results of 50 pCil gross alpha and 55 pCil gross beta and it was analyzed for isotopic uranium Ra-226 and Ra-228 The results were 74 pCil U-234 025 pCil U-235 037 pCil U-238 12 pCil Ra-226 and 7 pCil Ra-228 These values are from Table 10 of the report and it appears that the Ra-226 value was incorrectly transcribed in the CREW Technical Note as 37 pCil The isotopic uranium results do not match those expected as U-234 is elevated and results for U-235 and U-238 are almost equal Therefore the analytical results for this sample are questionable Furthermore all the rest of the ground water samples had gross alpha results of 3 pCil or less indicating that there are no significant amounts of radium in groundwater

In addition a limited literature review was performed to assess the potential for leaching of radium from subsurface solids by nitrate at the NMI Site The literature demonstrates that radium may be found in a handful of aqueous phase complexes The predominant species of radium in aqueous solution is the divalent cation Ra2+ but other complexes such as RaOH+ RaCl+ RaCO3 and RaSO4 may also be found (Beneš et al 1982 Beneš 1984) Radium appears to be the least likely of all the alkali earth metals (beryllium magnesium calcium strontium barium and radium) to form aqueous phase

11

de maximis inc

complexes (Kirby and Salutsky 1964) As such we are unclear as to the origin of Dr Landreacutersquos statement that ldquonitrates are known to influence greatly radium leaching from the soil into the groundwaterrdquo (Landreacute 2006) The salt Ra(NO3)2 is highly soluble (245 g100 g H2O Kirby and Salutsky 1964) but we have not found evidence to suggest that the species Ra(NO3)+ or Ra(NO3)2 exist in aqueous solution Lauria et al (2004) present radionuclide and water chemistry data from groundwater samples taken from the Buena Lagoon in Brazil A plot of Ra-226 and Ra-228 concentrations versus nitrate concentration is presented in Figure 1 These data demonstrate that both Ra-226 and Ra-228 are almost completely uncorrelated with nitrate at this site as evidenced by correlation coefficients (r2) of approximately 005 and 004 respectively Szabo et al (1997) also report that no correlation exists between nitrate and Ra-226 in 17 groundwater samples taken from the Kirkwood-Cohansey aquifer system in New Jersey although Ra-228 appears to be weakly correlated with nitrate in the same data set (r2 = 018)

Figure 1 Correlation between aqueous radium and nitrate (Lauria et al 2004)

r 2 = 005

r 2 = 004

1E-10 1E-09 1E-08 1E-07 1E-06 1E-05 1E-04

0 5 10 15 20 25

Nitrate (mgL)

Ra

(microg

L)

Ra-226 Ra-228 Linear (Ra-226) Linear (Ra-228)

The mobility of radium in the subsurface appears to be controlled primarily by sorption and precipitation (Lauria et al 2004) High values of aqueous phase radium have been observed in groundwater samples with either increased salinity or reduced pH (Kraemer and Reed 1984 Lauria and Godoy 2002) Demir (1989) demonstrated the increased mobility of radium in acidic environments due to the reduced sorptive capacity of sediments at low pH Figure 2 is a plot of radium concentration versus sample pH from the same data set (Lauria et al 2004)

12

de maximis inc

Figure 2 Correlation between aqueous radium and pH (Lauria et al 2004) 226Ra = -4E-06pH + 3E-05

r2 = 058

228Ra = -5E-08pH + 4E-07 r2 = 032

1E-10 1E-09 1E-08 1E-07 1E-06 1E-05 1E-04

3 4 5 6 7

pH

Ra

(microg

L)

Ra-226 Ra-228 Linear (Ra-226) Linear (Ra-228)

Here a stronger correlation between Ra-226 concentration and pH exists (r2 = 058) along with an observable trend of decreasing aqueous phase radium with increasing pH Szabo et al (1997) also report a negative correlation (r2 = 030) between Ra-228 and pH but no significant correlation between Ra-226 and pH Divalent cations such as calcium and magnesium compete with radium for a limited number of surface complexation and adsorption sites (Szabo et al 1997) As lime (a mixture potentially containing calcium oxide calcium hydroxide and calcium carbonate) was co-disposed with uranium and nitric acid at the Holding Basin nitrate may serve as a site-specific tracer for conditions that may lead to the release of adsorbed or surface-complexed radium from the solid phase although no comparable calcium ldquoplumerdquo has been defined unlike nitrate in groundwater Therefore we suggest that Dr Landreacutersquos statement could be qualified by stating that low pH values are known to influence the leaching of radium from aquifer solids However as discussed above we do not see a compelling reason to add radium to the analytical list at this time

Please contact me if you have any questions

Sincerely

Bruce Thompson

cc Respondents Bennie Underwood PE de maximis inc inc Nelson Walter PE MACTEC Inc Peter Zeeb PhD PG GeoSyntec Consultants Inc

13

de maximis inc

Figure 1 ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

Figure 2 Comparison of Alpha Spectroscopy and ICP-MS Uranium Analysis

Attachment 1 Metcalf amp Eddy Memorandum ndash Uranium Analysis comparison of alpha spectrometry and ICP-MS results

Attachment 2 QAPP SOPs for Alpha Spectroscopy and ICPMS Sample Preparation

Attachment 3 Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor Department of the Army January 19 2000

References

Beneš P Migration of radium in the terrestrial hydrosphere The behavior of radium in waterways and aquifers Technical Reports Series No 310 International Atomic Energy Agency IAEA-TECDOC-301 Vienna Austrai pp 119-173 1984

Beneš P M Obdržaacutelek M Čejchanovaacute The physicochemical forms of tracers of radium in aqueous solutions containing chlorides sulfates and carbonates Radiochem Radioanal Letters 50(4) pp 227-242 1982

CREW 2006 ldquoComments on Radiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Technical Note No 2006-01 January 6 2006

Demir I Temporal variations and sources of Pb Co Cu Ni Fe and Mo in shallow ground water of the McDonalds Branch basin Lebanon State Forest Burlington County New Jersey New Jersey Geological Survey Report 20 19 p 1989

EPA 2000 Guidance for Data Quality Assessment Practical Methods for Data Analysis EPA QAG-9 July 2000

EPA 2002 Guidance for Comparing Background and Chemical Concentrations in Soil for CERCLA Site EPA 540-R-01-003 September 2002

FSP 2004 Field Sampling Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Gilbert 1989 Statistical Methods for Environmental Pollution Monitoring Richard O Gilbert 1987

Kirby H W M L Salutsky The Radiochemistry of Radium National Academy of Sciences ndash National Research Council NAS-NS 3057 205 p 1964

14

de maximis inc

Kramer T F and D F Reid The occurrence and behavior of radium in saline formation water of the US Gulf Coast Region Isotope Geoscience 2 pp 153shy174 1984

Landreacute V Comments on ldquoRadiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Nuclear Metals Superfund Site CREW Technical Note 2006-01 January 6 2006

Lauria D C and J M O Godoy Abnormal high natural radium concentration in surface waters Journal of Environmental Radioactivity 61(2) pp 27-37 2002

Lauria D C R M R Almeida O Sracek Behavior of radium thorium and uranium in groundwater near the Buena Lagoon in the Coastal Zone of the State of Rio de Janeiro Brazil Environmental Geology 47 pp 11-19 200

MACTEC 2005 Radiological Review of Fall 2004 RI Analytical Data Nuclear Metals Superfund Site Concord Massachusetts September 26 2004

MampE 2005 Memorandum from Andrew Schkuta to Ed Conroy Subject Uranium Analysis comparison of alpha spectrometry and ICPMS results December 15 2005

PSOP Project Summary and Operations Plan Nuclear Metals Superfund Site Concord Massachusetts April 15 2005

QAPP 2004 Quality Assurance Project Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Szabo Z D E Rice C L MacLeod and T H Barringer Relation of distribution of radium nitrate and pesticides to agricultural land use and depth Kirkwood-Cohansey aquifer system New Jersey Coastal Plain 1990-1991 US Geological Survey Water-Resources Investigations Report 96-4165A 119 p 1997

15

bull

bull bull

bull bull

bull bull bull

bull bull

bull bull

bull

500

1500

Figure 1

Con

cent

ratio

n (m

gkg

)

ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

1000

0

0 5 10 15 20 25 30 35 40

Samples

Figure 2

Alp

ha S

pect

osco

py (m

gkg

)

Fall 2004 Total Uranium No Super-Depleted

1000

100

10

1

y = 10604x R2 = 09049

1 10 100 1000

ICPMS (mgkg)

Fall 2004 Super-Depleted Linear Regression

Attachment 1Connolly Jayme

From Taylor Mel issagepam ail epa gov Sent Wednesday December 21 2005 932AM To brucetdemaximiscom jhuntdemaximiscom Subject Fw Uranium Analysis Issues

Rad-versus-ICPMS RADvsiCPMS_soilse MEvsPRP_ICPMSP memodoc (158 dPDF (29 KB) OF (25 KB)

brucejohn here is mampes take on the rns vs alpha spec analysis discrepancy

Forwarded by MelissaG TaylorR1USEPAUS on 12212005 0930 AM

Conroy Ed ltEdConroyregrn-ea ecomcomgt To

MelissaG TaylorR1USEPAUSregEPA 12152005 0416 cc PM

Subject FW Uranium Analysis Issues

Melissa

Heres Andys take on the alpha spec vs ICPMS debate along with a couple of graphs showing correlation between the two methods and MampE split data vs PRP data

Please call me on the cell phone tomorrow to discuss

Ed

Edward A Conroy PE

Project Manager

METCALF amp EDDYJAECOM

7812246880 ph

6172832942 cell

1

ed conroym-e aecom com Attachment 1

-----Original Message----shyFrom Schkuta Andrew Sent Thursday December 15 2005 358 PM To Conroy Ed Subject Uranium Analysis Issues

Ed

Please find attached a memorandum that summarizes my review of why ICP-MS and alpha spec data may not necessarily agree

ltltRad-versus-ICPMS memodocgtgt

In addition I have attached 2 preliminary charts that have the following draft information

ltltRADvsiCPMS soilsedPDFgtgt

The RADvsiCPMS soilsed chart is a draft plot of MampEs alpha spec versus ICP-MS results when data was available for both samples It shows the same general trend as the PRP data with fairly good agreement between the two analyses There appears to be a slight positive bias for alpha spec data at low concentrations but this is expected as discussed in my memo

ltltMEvsPRP ICPMSPDFgtgt

The MEvsPRP ICPMS is a draft plot of MampE vs the PRPs total uranium results as measured by ICP-MS -With the exception of a few outliers which will be further investigated there appears to be good agreement between MampEs and the PRPs data As a note the median percent difference between MampEs and the PRPs data was 1

Andrew Schkuta CHMM

Technical Specialist - Chemistry

METCALF amp EDDYJAECOM

701 Edgewater Drive

Wakefield MA 01880

(781) 224-6353

(781) 245-6293 fax

2

(See attached file Rad-versus-ICPMS Attachment 1memodoc) (See attached file RADvsiCPMS soilsedPDF) (See attached file MEvsPRP ICPMSPDF)

3

Attachment 1

Metcalf amp Eddy Inc

METCALFampEDDY

Memorandum

Date December 15 2005

To Ed Conroy

From Andrew Schkuta

Subject Uranium Analysis comparison of alpha spectrometry and ICP-MS results

File WA151 Nuclear Metals Distribution

Summary This memorandum discusses the differences between uranium analysis conducted by two different methods alpha spectroscopy (alpha spec) and inductively coupled plasma shymass spectrometry (ICP-MS) In summary the alpha spec data is expected to be biased high compared to the ICP-MS data but this is not indicative that either method is being performed incorrectly

In the September 25 2005 technical memorandum Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts (MACTEC 2005) data was presented comparing alpha spec and ICP-MS analysis for uranium analysis and it was proposed that future uranium analysis be performed by ICP-MS only For samples analyzed that were collected through Spring 2005 all samples were analyzed by ICP-MS with a subset (20) analyzed by alpha spec It was noted that there appeared to be a positive bias in the alpha spec data compared to the ICP-MS data and some concern has been voiced if analysis by ICP-MS only would meet project objectives This memorandum provides background on why the two analytical methods may yield differing results and discusses how they may impact project quality objectives

There are two primary reasons why alpha spec is expected to yield results that may be positively biased

1 Sample Preparation The sample preparation method for alpha spec is more rigorous The preparation method in alpha spec breaks down the silica mineral structure of the sample By design the ICP-MS sample preparation does not break down the mineral structure and is intended to remove all metals that could become environmentally available

2 Sample Quanitation The results for alpha spec are adjusted for analyte loss by correcting the result based on the amount of tracer recovered (yield) ICP-MS data is absolute ie data is not corrected for loss during sample preparation

2 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

These two reasons are discussed in further detail as follows

Sample Preparation

Radionuclide analysis traditionally uses a sample preparation method that includes hydrofluoric acid (HF) to break down the sample so that all of the uranium present in the sample is put into solution On the other hand standard EPA methods for metals analysis including ICP-MS do not measure the total metals present in a sample For example method SW846-3050 (EPA 1996) (as specified by the PRP in their approved QAPP for metals analysis by ICP-MS) states

This method is not a total digestion technique for most samples It is a very strong acid digestion that will dissolve almost all elements that could become environmentally available By design elements bound in silicate structures are not normally dissolved by this procedure as they are not usually mobile in the environment If absolute total digestion is required use Method 3052

As noted by the method a SW846 method is available to do a total digestion (equivalent to the alpha spec method) as method 3052 However method 3052 (EPA 1996) states (emphasis added)

Note This technique is not appropriate for regulatory applications that require the use of leachate preparations (ie Method 3050 Method 3051 Method 1311 Method 1312 Method 1310 Method 1320 Method 1330 Method 3031 Method 3040) This method is appropriate for those applications requiring a total decomposition for research purposes (ie~ geological studiesr mass balancesr analysis of Standard Reference Materials) or in response to a regulation that requires total sample decomposition

Even though the ICP-MS preparation method does not yield total results it still may conservatively measure the amount of metals that are environmentally available In the Research Project Summary Ambient Levels of Metals in New Jersey Soils prepared by the New Jersey Department of Environmental Protection (NJDEP 2003) where samples were prepared using SW846-3050 procedures it states

Metal concentrations reported in this summary were acid-extractable metals not necessarily total metal concentrations The extraction method (USEPA Method 3050) is a vigorous extraction method designed to remove all metals that could become environmentally available In practice the extraction method likely overestimates metals that could become available since the sample is refluxed with both concentrated nitric acid and hydrogen peroxide However by design the method will not extract chemicals incorporated in silica minerals as they are usually not mobile in the environment Thus concentrations reported in these studies may be lower than those indicated from analyses using methods designed to measure total metal concentrations such as x-ray fluorescence methods

Although references to uranium were not located it is expected that uranium will have some of the same type of mobility characteristics as other metals in the soil matrix The paper Ground Water Issue Behavior of Metals in Soil published by EPAs Superfund Technology Support Center for Ground Water (EPA 1992) describes how metals are distributed within soil The paper describes how metals are partitioned in various pools and how a rigorous extraction using HF is required to release metals present in the structure of primary or secondary minerals It is

3 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

expected that native metals (ie background concentrations) can be present in these structures The paper indicates that for four metals examined (cadmium copper zinc and nickel) the amount of metals that was bound in the mineral structure ranges from approximately 15-60 It would be expected that naturally occurring uranium would have similar characteristics

The National Institutes of Standards amp Technology (NIST) recognizes the differences between total digestion and method SW846-3050 and has issued Addendum to Certificates for the Certificate of Analysis (NIST 2003) for some Standard Reference Materials (SRMs) The Addendum to Certificates for SRM 2711 a Montana agricultural soil indicates that trace metals were measured at approximately 40 to 90 of their certified value when analyzed using SW846-3050 as the preparation method Even though uranium is not listed as a target analyte it is reasonable to expect that is would behave in a similar fashion

Sample Quanitation

The preparation method for alpha spec analysis includes addition of a tracer (U-232) to sample and is carried through the entire analytical scheme The amount of tracer is measured during analysis and the results for target compounds (the other uranium isotopes) are corrected for the amount of tracer recovery (yield) For example if amount of tracer measured is 50 of what was initially added to the sample the sample results are all corrected by dividing by 050 In terms of method quality assurancequality control (QAQC) the PRP QAPP indicates that tracer recovery in the range of 15-125 is considered acceptable and the analytical system is considered in control In other words up to 85 of the compound can be lost during sample preparation and analysis and the data is considered meeting all QAQC objectives This method of analysis is the equivalent to the technique of isotope dilution which is used for some non-radiological environmental analyses (eg EPA Method 1625 Semivolatile Organic Compounds by Isotope Dilution GCMS) Analyses using this technique although considered the most accurate are not widely employed due to the complexity and expense of using the isotope dilution technique

ICP-MS analysis measures the concentration of analyte with no correction for potential losses However QAQC procedures are in place to make sure the losses are not considered significant In addition to laboratory control samples matrix spike analysis measures the recovery for the target analyte that was over-spiked into the sample The PRP QAPP states that matrix spike recovery should be within the range of 75-125 Therefore up to 25 of the analyte could be lost and the method has still met the QAQC objectives If recovery is less than this then validation procedures acknowledge that the data may be biased low and sample results are flagged as estimated

Therefore because the alpha spec data is corrected for analyte loss during the analytical process while ICP-MS data is not some amount of positive bias is expected to be present in alpha spec data when compared to ICP-MS data

Considerations

In evaluating which analysis is appropriate for samples from the project neither method is inherently better than the other The following factors should be considered

1 What is being measured ICP-MS measures mass directly and data can be mathematically converted to activity Alpha spec measures activity directly and can be mathematically converted to mass

4 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

2 What are the objectives of the data in terms of measuring total versus environmentally available The ICP-MS method specified in the PRP QAPP measures uranium concentrations in the same manner that EPA has measured other metals concentrations Should uranium mass be measured in a typical way that will be used for measuring other metals of concern at the site (eg copper beryllium arsenic etc)

3 If ICP-MS is more cost effective that alpha spec should the option of using preparation method SW846-3052 be considered This should yield total uranium data similar to alpha spec analysis while still using the efficiencies of ICP-MS analysis

4 What units will the cleanup level for the site be in If its in mass units then ICPshyMS analysis may be more appropriate If its in activity units then alpha spec analysis may be warranted

References

Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts Technical Memorandum from Heath Downey to Bruce Thompson de maxim is inc September 26 2005 (MACTEC 2005)

Test Methods for Evaluating Solid Waste PhysicalChemical Methods SW-846 Third Edition with updates EPA 1996

Ambient Levels of Metals in New Jersey Soils Environmental Assessment and Risk Analysis Element Research Project Summary Division of Science Research and Technology State of New Jersey Department of Environmental Protection May2003

Ground Water Issue Behavior of Metals in Soil EPA Superfund Technology Support Center for Ground Water EPA540S-92018 October 1993

Certificate of Analysis Standard Reference Material 2711 Montana Soil National Institute of Standards amp Technology July 18 2003

bull bull

bull bull bullbull bull

10000

MampE SoilSediment Results Total Uranium mgKg

bull

bull - ~-~-----~~---~~-~--~~1000

en aE

I I bull Total a u

I -

bull bull 100 1shy -middot- -middot bullbull bull

Ybull bull

10 100 1000 10000

Alpha Spec

Attachm

ent 1

10 -

100000

TOTAL URANIUM BY ICP-MS mgKg

bull

bull 10000

I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
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                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
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                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
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                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 9: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

de maximis inc

source of the uranium (whether depleted or natural uranium) is not relevant For risk assessment purposes all uranium will be considered and the contribution to risk associated with background constituents of concern will be considered after the total risk is calculated

Measurements of environmental data rarely have a normal (Gaussian) distribution and typically are lognormal distributions (Gilbert 1987) EPA guidance indicates that ldquoit is unusual to encounter environmental data sets that are normally distributedrdquo (EPA 2002) and ldquoenvironmental data commonly exhibit frequency distributions that are non-negative and skewed with heavy or long right tailsrdquo (EPA 2000) The composition of uranium (natural or depleted) will be used during risk assessment but it has limited usefulness otherwise However the isotopic composition of uranium will be one of the issues considered during data evaluation to avoid the sort of issues pointed out in the CREW Technical Note

Response to Section 4 - Thorium evaluation

de maximis inc re-iterates that the Technical Memorandum shows thorium-232 and thorium-228 concentrations are roughly equal (and considered within the analytical variability for alpha spectroscopy) and that thorium-230 at the Site is not related to Site activities Furthermore uranium-232 (from recycled uranium) is not present at the site as it is primarily a concern with enriched uranium Even if present it would not be in significant enough quantities to impact thorium-228 concentrations

Refer to previous discussion pertaining to ICPMS versus alpha spectroscopy analysis information Also note that ICPMS instrument sensitivity for thorium is very high due to the high mass to charge ratio (ie it is a good responder) Naturally occurring thorium is 100 Th-232 No data quality issues were observed during data validation and data has been deemed usable

Lastly thorium was used at the Site but at significantly lower amounts when compared to uranium Therefore low concentrations (within background concentration range) of thorium are not unexpected and historical data have demonstrated this

Response to Section 5 - Other radionuclides

There is no way to know the exact pedigree of the depleted uranium processed at the Site or to determine which portion of the depleted uranium might have come from reprocessed fuel rods and therefore contain small amounts of reactor generated radionuclides Studies of depleted uranium have shown that byproduct radionuclides in uranium are present at very low levels In addition dose assessments of these trace contaminants have shown that the radiological dose from these radionuclides is insignificant compared to that from uranium These conclusions were documented in a January 2000 Department of the Army analysis of transuranics and other contaminants in depleted uranium armor (Attachment 3) Of particular relevance is the fact that the Department of the Army analysis took samples of DU billets from Starmet for this study The Army concluded that ldquoThe levels of transuranics elements (TRUs) and fission

9

de maximis inc

products (FP) are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is saferdquo If the TRUs and FP contribute less than a one percent increase in dose for DU metal then they would contribute an even more negligible dose at the 10s to 1000s of mgkg of DU present in NMI Site media With respect to the HTDR analysis during the RI historical data indicated that the holding basin and former drain outfall regions of the bog were the most likely areas to have significant concentrations of uranium therefore the RI work plan identified locations to be sampled in these areas (FSP 2004) Further the approach outlined in the RI was to evaluate for HTDR in samples with elevated concentrations of uranium not necessarily the highest concentrations There were other AOIs where uranium was detected in elevated concentrations but no sample was collected for HTDR However since the holding basin was in use for the vast majority of time that uranium was actively processed at the site and no HTDR were detected in the six samples from this area it is unlikely that HTDR would be identified in other areas of the site

Samples for HTDRs were collected according to the RI (FSP 2004) such that 3 of the samples from AOI 1 were above the water table and 3 of the samples were from below the water table one sample was collected from each of the AOI 6 former drain outfall areas of the bog and one sample collected from the buried drum excavation (AOI 2) These HTDR samples are representative for the areas investigated (AOI 1 AOI 2 and AOI 6) and of the site based on the distribution of total uranium results described below

bull Total uranium in the HTDR sample locations from AOI 1 ranged from 38 to 511 mgkg while total uranium for all samples in AOI 1 ranged from 11 to 672 mgkg Only 25 samples from all samples in AOI 1 had total uranium concentrations greater than 100 mgkg and 4 of the 6 HTDR samples were from this subset

bull Total uranium in the HTDR sample locations from AOI 6 were 25 mgkg at former drain outfall 1 and 4550 mgkg at drain outfall 2 while total uranium for all samples in AOI 6 ranged from 16 to 4550 mgkg Only 10 samples from all samples in AOI 6 had total uranium concentrations greater than 100 mgkg and one HTDR sample was from the maximum total uranium sample location

bull The HTDR sample from AOI 2 drum excavation bottom had a total uranium concentration of 18 mgkg while total uranium for all excavation sides and bottom ranged from 14 to 49 mgkg

Only a few samples have uranium concentrations greater than 1000 mgkg and 6 of the 8 are from samples deemed as lsquosuper-depletedrsquo These 6 locations will be re-sampled as part of Phase 1C since the ICPMS analysis results are considered an anomaly Sporadic low detections of HTDRs do not necessarily indicate the presence of HTDRs as false positive results are common for these analyses Although some HTDRs have been detected in the past they are sporadic (not all samples have detections) and do not indicate the presence of HTDRs at significant levels

10

de maximis inc

It is also important to note that the historical data provided in the RI (PSOP 2005) was used to assist in design of the RI (FSP 2004) Data were compiled from many sources of information and represent samples collected analyzed and reported using various sample collection procedures laboratory analytical methods and data reporting conventions Data compilation activities involved both manual hand-entries and electronic entries when applicable Based on a review of the electronic database severe data quality limitations were noted including unknown analytical quality and validation levels missing location coordinates inconsistent media codes unknown detection limits and radionuclide uncertainty measurements inconsistent andor incorrect reporting of radionuclides using mass-based and activity-based units These limitations were considered significant during the scoping of the RIFS Work Plan therefore the historical data were used primarily to evaluate the spatial distribution of parameters and magnitude of detected concentrations The limitations of the historical data were used to provide the basis for identifying data gaps and scoping the RIFS As such the FSP included analysis for HTDRs at the areas of the Site known to be impacted by historical practices

With respect to comments provided on U-232 and U-236 these uranium isotopes are produced during the fission of uranium (as are plutonium americium and the rest of the HTDRs) Since none of these other fission products have been detected at the Site it is not credible that either uranium-232 or uranium-236 would be present

Response to Section 6 - Radium

The historical radium in groundwater data is suspect for several reasons including the previous discussion on historical data provided in the RI (PSOP 2005) Groundwater samples were collected from 13 wells at the Site (single sampling event) using bailers as part of this investigation All samples were screened for radioactivity by gross alpha beta analysis Only one sample exceeded the action level (gt 10 pCil gross alpha) with results of 50 pCil gross alpha and 55 pCil gross beta and it was analyzed for isotopic uranium Ra-226 and Ra-228 The results were 74 pCil U-234 025 pCil U-235 037 pCil U-238 12 pCil Ra-226 and 7 pCil Ra-228 These values are from Table 10 of the report and it appears that the Ra-226 value was incorrectly transcribed in the CREW Technical Note as 37 pCil The isotopic uranium results do not match those expected as U-234 is elevated and results for U-235 and U-238 are almost equal Therefore the analytical results for this sample are questionable Furthermore all the rest of the ground water samples had gross alpha results of 3 pCil or less indicating that there are no significant amounts of radium in groundwater

In addition a limited literature review was performed to assess the potential for leaching of radium from subsurface solids by nitrate at the NMI Site The literature demonstrates that radium may be found in a handful of aqueous phase complexes The predominant species of radium in aqueous solution is the divalent cation Ra2+ but other complexes such as RaOH+ RaCl+ RaCO3 and RaSO4 may also be found (Beneš et al 1982 Beneš 1984) Radium appears to be the least likely of all the alkali earth metals (beryllium magnesium calcium strontium barium and radium) to form aqueous phase

11

de maximis inc

complexes (Kirby and Salutsky 1964) As such we are unclear as to the origin of Dr Landreacutersquos statement that ldquonitrates are known to influence greatly radium leaching from the soil into the groundwaterrdquo (Landreacute 2006) The salt Ra(NO3)2 is highly soluble (245 g100 g H2O Kirby and Salutsky 1964) but we have not found evidence to suggest that the species Ra(NO3)+ or Ra(NO3)2 exist in aqueous solution Lauria et al (2004) present radionuclide and water chemistry data from groundwater samples taken from the Buena Lagoon in Brazil A plot of Ra-226 and Ra-228 concentrations versus nitrate concentration is presented in Figure 1 These data demonstrate that both Ra-226 and Ra-228 are almost completely uncorrelated with nitrate at this site as evidenced by correlation coefficients (r2) of approximately 005 and 004 respectively Szabo et al (1997) also report that no correlation exists between nitrate and Ra-226 in 17 groundwater samples taken from the Kirkwood-Cohansey aquifer system in New Jersey although Ra-228 appears to be weakly correlated with nitrate in the same data set (r2 = 018)

Figure 1 Correlation between aqueous radium and nitrate (Lauria et al 2004)

r 2 = 005

r 2 = 004

1E-10 1E-09 1E-08 1E-07 1E-06 1E-05 1E-04

0 5 10 15 20 25

Nitrate (mgL)

Ra

(microg

L)

Ra-226 Ra-228 Linear (Ra-226) Linear (Ra-228)

The mobility of radium in the subsurface appears to be controlled primarily by sorption and precipitation (Lauria et al 2004) High values of aqueous phase radium have been observed in groundwater samples with either increased salinity or reduced pH (Kraemer and Reed 1984 Lauria and Godoy 2002) Demir (1989) demonstrated the increased mobility of radium in acidic environments due to the reduced sorptive capacity of sediments at low pH Figure 2 is a plot of radium concentration versus sample pH from the same data set (Lauria et al 2004)

12

de maximis inc

Figure 2 Correlation between aqueous radium and pH (Lauria et al 2004) 226Ra = -4E-06pH + 3E-05

r2 = 058

228Ra = -5E-08pH + 4E-07 r2 = 032

1E-10 1E-09 1E-08 1E-07 1E-06 1E-05 1E-04

3 4 5 6 7

pH

Ra

(microg

L)

Ra-226 Ra-228 Linear (Ra-226) Linear (Ra-228)

Here a stronger correlation between Ra-226 concentration and pH exists (r2 = 058) along with an observable trend of decreasing aqueous phase radium with increasing pH Szabo et al (1997) also report a negative correlation (r2 = 030) between Ra-228 and pH but no significant correlation between Ra-226 and pH Divalent cations such as calcium and magnesium compete with radium for a limited number of surface complexation and adsorption sites (Szabo et al 1997) As lime (a mixture potentially containing calcium oxide calcium hydroxide and calcium carbonate) was co-disposed with uranium and nitric acid at the Holding Basin nitrate may serve as a site-specific tracer for conditions that may lead to the release of adsorbed or surface-complexed radium from the solid phase although no comparable calcium ldquoplumerdquo has been defined unlike nitrate in groundwater Therefore we suggest that Dr Landreacutersquos statement could be qualified by stating that low pH values are known to influence the leaching of radium from aquifer solids However as discussed above we do not see a compelling reason to add radium to the analytical list at this time

Please contact me if you have any questions

Sincerely

Bruce Thompson

cc Respondents Bennie Underwood PE de maximis inc inc Nelson Walter PE MACTEC Inc Peter Zeeb PhD PG GeoSyntec Consultants Inc

13

de maximis inc

Figure 1 ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

Figure 2 Comparison of Alpha Spectroscopy and ICP-MS Uranium Analysis

Attachment 1 Metcalf amp Eddy Memorandum ndash Uranium Analysis comparison of alpha spectrometry and ICP-MS results

Attachment 2 QAPP SOPs for Alpha Spectroscopy and ICPMS Sample Preparation

Attachment 3 Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor Department of the Army January 19 2000

References

Beneš P Migration of radium in the terrestrial hydrosphere The behavior of radium in waterways and aquifers Technical Reports Series No 310 International Atomic Energy Agency IAEA-TECDOC-301 Vienna Austrai pp 119-173 1984

Beneš P M Obdržaacutelek M Čejchanovaacute The physicochemical forms of tracers of radium in aqueous solutions containing chlorides sulfates and carbonates Radiochem Radioanal Letters 50(4) pp 227-242 1982

CREW 2006 ldquoComments on Radiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Technical Note No 2006-01 January 6 2006

Demir I Temporal variations and sources of Pb Co Cu Ni Fe and Mo in shallow ground water of the McDonalds Branch basin Lebanon State Forest Burlington County New Jersey New Jersey Geological Survey Report 20 19 p 1989

EPA 2000 Guidance for Data Quality Assessment Practical Methods for Data Analysis EPA QAG-9 July 2000

EPA 2002 Guidance for Comparing Background and Chemical Concentrations in Soil for CERCLA Site EPA 540-R-01-003 September 2002

FSP 2004 Field Sampling Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Gilbert 1989 Statistical Methods for Environmental Pollution Monitoring Richard O Gilbert 1987

Kirby H W M L Salutsky The Radiochemistry of Radium National Academy of Sciences ndash National Research Council NAS-NS 3057 205 p 1964

14

de maximis inc

Kramer T F and D F Reid The occurrence and behavior of radium in saline formation water of the US Gulf Coast Region Isotope Geoscience 2 pp 153shy174 1984

Landreacute V Comments on ldquoRadiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Nuclear Metals Superfund Site CREW Technical Note 2006-01 January 6 2006

Lauria D C and J M O Godoy Abnormal high natural radium concentration in surface waters Journal of Environmental Radioactivity 61(2) pp 27-37 2002

Lauria D C R M R Almeida O Sracek Behavior of radium thorium and uranium in groundwater near the Buena Lagoon in the Coastal Zone of the State of Rio de Janeiro Brazil Environmental Geology 47 pp 11-19 200

MACTEC 2005 Radiological Review of Fall 2004 RI Analytical Data Nuclear Metals Superfund Site Concord Massachusetts September 26 2004

MampE 2005 Memorandum from Andrew Schkuta to Ed Conroy Subject Uranium Analysis comparison of alpha spectrometry and ICPMS results December 15 2005

PSOP Project Summary and Operations Plan Nuclear Metals Superfund Site Concord Massachusetts April 15 2005

QAPP 2004 Quality Assurance Project Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Szabo Z D E Rice C L MacLeod and T H Barringer Relation of distribution of radium nitrate and pesticides to agricultural land use and depth Kirkwood-Cohansey aquifer system New Jersey Coastal Plain 1990-1991 US Geological Survey Water-Resources Investigations Report 96-4165A 119 p 1997

15

bull

bull bull

bull bull

bull bull bull

bull bull

bull bull

bull

500

1500

Figure 1

Con

cent

ratio

n (m

gkg

)

ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

1000

0

0 5 10 15 20 25 30 35 40

Samples

Figure 2

Alp

ha S

pect

osco

py (m

gkg

)

Fall 2004 Total Uranium No Super-Depleted

1000

100

10

1

y = 10604x R2 = 09049

1 10 100 1000

ICPMS (mgkg)

Fall 2004 Super-Depleted Linear Regression

Attachment 1Connolly Jayme

From Taylor Mel issagepam ail epa gov Sent Wednesday December 21 2005 932AM To brucetdemaximiscom jhuntdemaximiscom Subject Fw Uranium Analysis Issues

Rad-versus-ICPMS RADvsiCPMS_soilse MEvsPRP_ICPMSP memodoc (158 dPDF (29 KB) OF (25 KB)

brucejohn here is mampes take on the rns vs alpha spec analysis discrepancy

Forwarded by MelissaG TaylorR1USEPAUS on 12212005 0930 AM

Conroy Ed ltEdConroyregrn-ea ecomcomgt To

MelissaG TaylorR1USEPAUSregEPA 12152005 0416 cc PM

Subject FW Uranium Analysis Issues

Melissa

Heres Andys take on the alpha spec vs ICPMS debate along with a couple of graphs showing correlation between the two methods and MampE split data vs PRP data

Please call me on the cell phone tomorrow to discuss

Ed

Edward A Conroy PE

Project Manager

METCALF amp EDDYJAECOM

7812246880 ph

6172832942 cell

1

ed conroym-e aecom com Attachment 1

-----Original Message----shyFrom Schkuta Andrew Sent Thursday December 15 2005 358 PM To Conroy Ed Subject Uranium Analysis Issues

Ed

Please find attached a memorandum that summarizes my review of why ICP-MS and alpha spec data may not necessarily agree

ltltRad-versus-ICPMS memodocgtgt

In addition I have attached 2 preliminary charts that have the following draft information

ltltRADvsiCPMS soilsedPDFgtgt

The RADvsiCPMS soilsed chart is a draft plot of MampEs alpha spec versus ICP-MS results when data was available for both samples It shows the same general trend as the PRP data with fairly good agreement between the two analyses There appears to be a slight positive bias for alpha spec data at low concentrations but this is expected as discussed in my memo

ltltMEvsPRP ICPMSPDFgtgt

The MEvsPRP ICPMS is a draft plot of MampE vs the PRPs total uranium results as measured by ICP-MS -With the exception of a few outliers which will be further investigated there appears to be good agreement between MampEs and the PRPs data As a note the median percent difference between MampEs and the PRPs data was 1

Andrew Schkuta CHMM

Technical Specialist - Chemistry

METCALF amp EDDYJAECOM

701 Edgewater Drive

Wakefield MA 01880

(781) 224-6353

(781) 245-6293 fax

2

(See attached file Rad-versus-ICPMS Attachment 1memodoc) (See attached file RADvsiCPMS soilsedPDF) (See attached file MEvsPRP ICPMSPDF)

3

Attachment 1

Metcalf amp Eddy Inc

METCALFampEDDY

Memorandum

Date December 15 2005

To Ed Conroy

From Andrew Schkuta

Subject Uranium Analysis comparison of alpha spectrometry and ICP-MS results

File WA151 Nuclear Metals Distribution

Summary This memorandum discusses the differences between uranium analysis conducted by two different methods alpha spectroscopy (alpha spec) and inductively coupled plasma shymass spectrometry (ICP-MS) In summary the alpha spec data is expected to be biased high compared to the ICP-MS data but this is not indicative that either method is being performed incorrectly

In the September 25 2005 technical memorandum Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts (MACTEC 2005) data was presented comparing alpha spec and ICP-MS analysis for uranium analysis and it was proposed that future uranium analysis be performed by ICP-MS only For samples analyzed that were collected through Spring 2005 all samples were analyzed by ICP-MS with a subset (20) analyzed by alpha spec It was noted that there appeared to be a positive bias in the alpha spec data compared to the ICP-MS data and some concern has been voiced if analysis by ICP-MS only would meet project objectives This memorandum provides background on why the two analytical methods may yield differing results and discusses how they may impact project quality objectives

There are two primary reasons why alpha spec is expected to yield results that may be positively biased

1 Sample Preparation The sample preparation method for alpha spec is more rigorous The preparation method in alpha spec breaks down the silica mineral structure of the sample By design the ICP-MS sample preparation does not break down the mineral structure and is intended to remove all metals that could become environmentally available

2 Sample Quanitation The results for alpha spec are adjusted for analyte loss by correcting the result based on the amount of tracer recovered (yield) ICP-MS data is absolute ie data is not corrected for loss during sample preparation

2 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

These two reasons are discussed in further detail as follows

Sample Preparation

Radionuclide analysis traditionally uses a sample preparation method that includes hydrofluoric acid (HF) to break down the sample so that all of the uranium present in the sample is put into solution On the other hand standard EPA methods for metals analysis including ICP-MS do not measure the total metals present in a sample For example method SW846-3050 (EPA 1996) (as specified by the PRP in their approved QAPP for metals analysis by ICP-MS) states

This method is not a total digestion technique for most samples It is a very strong acid digestion that will dissolve almost all elements that could become environmentally available By design elements bound in silicate structures are not normally dissolved by this procedure as they are not usually mobile in the environment If absolute total digestion is required use Method 3052

As noted by the method a SW846 method is available to do a total digestion (equivalent to the alpha spec method) as method 3052 However method 3052 (EPA 1996) states (emphasis added)

Note This technique is not appropriate for regulatory applications that require the use of leachate preparations (ie Method 3050 Method 3051 Method 1311 Method 1312 Method 1310 Method 1320 Method 1330 Method 3031 Method 3040) This method is appropriate for those applications requiring a total decomposition for research purposes (ie~ geological studiesr mass balancesr analysis of Standard Reference Materials) or in response to a regulation that requires total sample decomposition

Even though the ICP-MS preparation method does not yield total results it still may conservatively measure the amount of metals that are environmentally available In the Research Project Summary Ambient Levels of Metals in New Jersey Soils prepared by the New Jersey Department of Environmental Protection (NJDEP 2003) where samples were prepared using SW846-3050 procedures it states

Metal concentrations reported in this summary were acid-extractable metals not necessarily total metal concentrations The extraction method (USEPA Method 3050) is a vigorous extraction method designed to remove all metals that could become environmentally available In practice the extraction method likely overestimates metals that could become available since the sample is refluxed with both concentrated nitric acid and hydrogen peroxide However by design the method will not extract chemicals incorporated in silica minerals as they are usually not mobile in the environment Thus concentrations reported in these studies may be lower than those indicated from analyses using methods designed to measure total metal concentrations such as x-ray fluorescence methods

Although references to uranium were not located it is expected that uranium will have some of the same type of mobility characteristics as other metals in the soil matrix The paper Ground Water Issue Behavior of Metals in Soil published by EPAs Superfund Technology Support Center for Ground Water (EPA 1992) describes how metals are distributed within soil The paper describes how metals are partitioned in various pools and how a rigorous extraction using HF is required to release metals present in the structure of primary or secondary minerals It is

3 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

expected that native metals (ie background concentrations) can be present in these structures The paper indicates that for four metals examined (cadmium copper zinc and nickel) the amount of metals that was bound in the mineral structure ranges from approximately 15-60 It would be expected that naturally occurring uranium would have similar characteristics

The National Institutes of Standards amp Technology (NIST) recognizes the differences between total digestion and method SW846-3050 and has issued Addendum to Certificates for the Certificate of Analysis (NIST 2003) for some Standard Reference Materials (SRMs) The Addendum to Certificates for SRM 2711 a Montana agricultural soil indicates that trace metals were measured at approximately 40 to 90 of their certified value when analyzed using SW846-3050 as the preparation method Even though uranium is not listed as a target analyte it is reasonable to expect that is would behave in a similar fashion

Sample Quanitation

The preparation method for alpha spec analysis includes addition of a tracer (U-232) to sample and is carried through the entire analytical scheme The amount of tracer is measured during analysis and the results for target compounds (the other uranium isotopes) are corrected for the amount of tracer recovery (yield) For example if amount of tracer measured is 50 of what was initially added to the sample the sample results are all corrected by dividing by 050 In terms of method quality assurancequality control (QAQC) the PRP QAPP indicates that tracer recovery in the range of 15-125 is considered acceptable and the analytical system is considered in control In other words up to 85 of the compound can be lost during sample preparation and analysis and the data is considered meeting all QAQC objectives This method of analysis is the equivalent to the technique of isotope dilution which is used for some non-radiological environmental analyses (eg EPA Method 1625 Semivolatile Organic Compounds by Isotope Dilution GCMS) Analyses using this technique although considered the most accurate are not widely employed due to the complexity and expense of using the isotope dilution technique

ICP-MS analysis measures the concentration of analyte with no correction for potential losses However QAQC procedures are in place to make sure the losses are not considered significant In addition to laboratory control samples matrix spike analysis measures the recovery for the target analyte that was over-spiked into the sample The PRP QAPP states that matrix spike recovery should be within the range of 75-125 Therefore up to 25 of the analyte could be lost and the method has still met the QAQC objectives If recovery is less than this then validation procedures acknowledge that the data may be biased low and sample results are flagged as estimated

Therefore because the alpha spec data is corrected for analyte loss during the analytical process while ICP-MS data is not some amount of positive bias is expected to be present in alpha spec data when compared to ICP-MS data

Considerations

In evaluating which analysis is appropriate for samples from the project neither method is inherently better than the other The following factors should be considered

1 What is being measured ICP-MS measures mass directly and data can be mathematically converted to activity Alpha spec measures activity directly and can be mathematically converted to mass

4 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

2 What are the objectives of the data in terms of measuring total versus environmentally available The ICP-MS method specified in the PRP QAPP measures uranium concentrations in the same manner that EPA has measured other metals concentrations Should uranium mass be measured in a typical way that will be used for measuring other metals of concern at the site (eg copper beryllium arsenic etc)

3 If ICP-MS is more cost effective that alpha spec should the option of using preparation method SW846-3052 be considered This should yield total uranium data similar to alpha spec analysis while still using the efficiencies of ICP-MS analysis

4 What units will the cleanup level for the site be in If its in mass units then ICPshyMS analysis may be more appropriate If its in activity units then alpha spec analysis may be warranted

References

Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts Technical Memorandum from Heath Downey to Bruce Thompson de maxim is inc September 26 2005 (MACTEC 2005)

Test Methods for Evaluating Solid Waste PhysicalChemical Methods SW-846 Third Edition with updates EPA 1996

Ambient Levels of Metals in New Jersey Soils Environmental Assessment and Risk Analysis Element Research Project Summary Division of Science Research and Technology State of New Jersey Department of Environmental Protection May2003

Ground Water Issue Behavior of Metals in Soil EPA Superfund Technology Support Center for Ground Water EPA540S-92018 October 1993

Certificate of Analysis Standard Reference Material 2711 Montana Soil National Institute of Standards amp Technology July 18 2003

bull bull

bull bull bullbull bull

10000

MampE SoilSediment Results Total Uranium mgKg

bull

bull - ~-~-----~~---~~-~--~~1000

en aE

I I bull Total a u

I -

bull bull 100 1shy -middot- -middot bullbull bull

Ybull bull

10 100 1000 10000

Alpha Spec

Attachm

ent 1

10 -

100000

TOTAL URANIUM BY ICP-MS mgKg

bull

bull 10000

I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 10: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

de maximis inc

products (FP) are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is saferdquo If the TRUs and FP contribute less than a one percent increase in dose for DU metal then they would contribute an even more negligible dose at the 10s to 1000s of mgkg of DU present in NMI Site media With respect to the HTDR analysis during the RI historical data indicated that the holding basin and former drain outfall regions of the bog were the most likely areas to have significant concentrations of uranium therefore the RI work plan identified locations to be sampled in these areas (FSP 2004) Further the approach outlined in the RI was to evaluate for HTDR in samples with elevated concentrations of uranium not necessarily the highest concentrations There were other AOIs where uranium was detected in elevated concentrations but no sample was collected for HTDR However since the holding basin was in use for the vast majority of time that uranium was actively processed at the site and no HTDR were detected in the six samples from this area it is unlikely that HTDR would be identified in other areas of the site

Samples for HTDRs were collected according to the RI (FSP 2004) such that 3 of the samples from AOI 1 were above the water table and 3 of the samples were from below the water table one sample was collected from each of the AOI 6 former drain outfall areas of the bog and one sample collected from the buried drum excavation (AOI 2) These HTDR samples are representative for the areas investigated (AOI 1 AOI 2 and AOI 6) and of the site based on the distribution of total uranium results described below

bull Total uranium in the HTDR sample locations from AOI 1 ranged from 38 to 511 mgkg while total uranium for all samples in AOI 1 ranged from 11 to 672 mgkg Only 25 samples from all samples in AOI 1 had total uranium concentrations greater than 100 mgkg and 4 of the 6 HTDR samples were from this subset

bull Total uranium in the HTDR sample locations from AOI 6 were 25 mgkg at former drain outfall 1 and 4550 mgkg at drain outfall 2 while total uranium for all samples in AOI 6 ranged from 16 to 4550 mgkg Only 10 samples from all samples in AOI 6 had total uranium concentrations greater than 100 mgkg and one HTDR sample was from the maximum total uranium sample location

bull The HTDR sample from AOI 2 drum excavation bottom had a total uranium concentration of 18 mgkg while total uranium for all excavation sides and bottom ranged from 14 to 49 mgkg

Only a few samples have uranium concentrations greater than 1000 mgkg and 6 of the 8 are from samples deemed as lsquosuper-depletedrsquo These 6 locations will be re-sampled as part of Phase 1C since the ICPMS analysis results are considered an anomaly Sporadic low detections of HTDRs do not necessarily indicate the presence of HTDRs as false positive results are common for these analyses Although some HTDRs have been detected in the past they are sporadic (not all samples have detections) and do not indicate the presence of HTDRs at significant levels

10

de maximis inc

It is also important to note that the historical data provided in the RI (PSOP 2005) was used to assist in design of the RI (FSP 2004) Data were compiled from many sources of information and represent samples collected analyzed and reported using various sample collection procedures laboratory analytical methods and data reporting conventions Data compilation activities involved both manual hand-entries and electronic entries when applicable Based on a review of the electronic database severe data quality limitations were noted including unknown analytical quality and validation levels missing location coordinates inconsistent media codes unknown detection limits and radionuclide uncertainty measurements inconsistent andor incorrect reporting of radionuclides using mass-based and activity-based units These limitations were considered significant during the scoping of the RIFS Work Plan therefore the historical data were used primarily to evaluate the spatial distribution of parameters and magnitude of detected concentrations The limitations of the historical data were used to provide the basis for identifying data gaps and scoping the RIFS As such the FSP included analysis for HTDRs at the areas of the Site known to be impacted by historical practices

With respect to comments provided on U-232 and U-236 these uranium isotopes are produced during the fission of uranium (as are plutonium americium and the rest of the HTDRs) Since none of these other fission products have been detected at the Site it is not credible that either uranium-232 or uranium-236 would be present

Response to Section 6 - Radium

The historical radium in groundwater data is suspect for several reasons including the previous discussion on historical data provided in the RI (PSOP 2005) Groundwater samples were collected from 13 wells at the Site (single sampling event) using bailers as part of this investigation All samples were screened for radioactivity by gross alpha beta analysis Only one sample exceeded the action level (gt 10 pCil gross alpha) with results of 50 pCil gross alpha and 55 pCil gross beta and it was analyzed for isotopic uranium Ra-226 and Ra-228 The results were 74 pCil U-234 025 pCil U-235 037 pCil U-238 12 pCil Ra-226 and 7 pCil Ra-228 These values are from Table 10 of the report and it appears that the Ra-226 value was incorrectly transcribed in the CREW Technical Note as 37 pCil The isotopic uranium results do not match those expected as U-234 is elevated and results for U-235 and U-238 are almost equal Therefore the analytical results for this sample are questionable Furthermore all the rest of the ground water samples had gross alpha results of 3 pCil or less indicating that there are no significant amounts of radium in groundwater

In addition a limited literature review was performed to assess the potential for leaching of radium from subsurface solids by nitrate at the NMI Site The literature demonstrates that radium may be found in a handful of aqueous phase complexes The predominant species of radium in aqueous solution is the divalent cation Ra2+ but other complexes such as RaOH+ RaCl+ RaCO3 and RaSO4 may also be found (Beneš et al 1982 Beneš 1984) Radium appears to be the least likely of all the alkali earth metals (beryllium magnesium calcium strontium barium and radium) to form aqueous phase

11

de maximis inc

complexes (Kirby and Salutsky 1964) As such we are unclear as to the origin of Dr Landreacutersquos statement that ldquonitrates are known to influence greatly radium leaching from the soil into the groundwaterrdquo (Landreacute 2006) The salt Ra(NO3)2 is highly soluble (245 g100 g H2O Kirby and Salutsky 1964) but we have not found evidence to suggest that the species Ra(NO3)+ or Ra(NO3)2 exist in aqueous solution Lauria et al (2004) present radionuclide and water chemistry data from groundwater samples taken from the Buena Lagoon in Brazil A plot of Ra-226 and Ra-228 concentrations versus nitrate concentration is presented in Figure 1 These data demonstrate that both Ra-226 and Ra-228 are almost completely uncorrelated with nitrate at this site as evidenced by correlation coefficients (r2) of approximately 005 and 004 respectively Szabo et al (1997) also report that no correlation exists between nitrate and Ra-226 in 17 groundwater samples taken from the Kirkwood-Cohansey aquifer system in New Jersey although Ra-228 appears to be weakly correlated with nitrate in the same data set (r2 = 018)

Figure 1 Correlation between aqueous radium and nitrate (Lauria et al 2004)

r 2 = 005

r 2 = 004

1E-10 1E-09 1E-08 1E-07 1E-06 1E-05 1E-04

0 5 10 15 20 25

Nitrate (mgL)

Ra

(microg

L)

Ra-226 Ra-228 Linear (Ra-226) Linear (Ra-228)

The mobility of radium in the subsurface appears to be controlled primarily by sorption and precipitation (Lauria et al 2004) High values of aqueous phase radium have been observed in groundwater samples with either increased salinity or reduced pH (Kraemer and Reed 1984 Lauria and Godoy 2002) Demir (1989) demonstrated the increased mobility of radium in acidic environments due to the reduced sorptive capacity of sediments at low pH Figure 2 is a plot of radium concentration versus sample pH from the same data set (Lauria et al 2004)

12

de maximis inc

Figure 2 Correlation between aqueous radium and pH (Lauria et al 2004) 226Ra = -4E-06pH + 3E-05

r2 = 058

228Ra = -5E-08pH + 4E-07 r2 = 032

1E-10 1E-09 1E-08 1E-07 1E-06 1E-05 1E-04

3 4 5 6 7

pH

Ra

(microg

L)

Ra-226 Ra-228 Linear (Ra-226) Linear (Ra-228)

Here a stronger correlation between Ra-226 concentration and pH exists (r2 = 058) along with an observable trend of decreasing aqueous phase radium with increasing pH Szabo et al (1997) also report a negative correlation (r2 = 030) between Ra-228 and pH but no significant correlation between Ra-226 and pH Divalent cations such as calcium and magnesium compete with radium for a limited number of surface complexation and adsorption sites (Szabo et al 1997) As lime (a mixture potentially containing calcium oxide calcium hydroxide and calcium carbonate) was co-disposed with uranium and nitric acid at the Holding Basin nitrate may serve as a site-specific tracer for conditions that may lead to the release of adsorbed or surface-complexed radium from the solid phase although no comparable calcium ldquoplumerdquo has been defined unlike nitrate in groundwater Therefore we suggest that Dr Landreacutersquos statement could be qualified by stating that low pH values are known to influence the leaching of radium from aquifer solids However as discussed above we do not see a compelling reason to add radium to the analytical list at this time

Please contact me if you have any questions

Sincerely

Bruce Thompson

cc Respondents Bennie Underwood PE de maximis inc inc Nelson Walter PE MACTEC Inc Peter Zeeb PhD PG GeoSyntec Consultants Inc

13

de maximis inc

Figure 1 ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

Figure 2 Comparison of Alpha Spectroscopy and ICP-MS Uranium Analysis

Attachment 1 Metcalf amp Eddy Memorandum ndash Uranium Analysis comparison of alpha spectrometry and ICP-MS results

Attachment 2 QAPP SOPs for Alpha Spectroscopy and ICPMS Sample Preparation

Attachment 3 Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor Department of the Army January 19 2000

References

Beneš P Migration of radium in the terrestrial hydrosphere The behavior of radium in waterways and aquifers Technical Reports Series No 310 International Atomic Energy Agency IAEA-TECDOC-301 Vienna Austrai pp 119-173 1984

Beneš P M Obdržaacutelek M Čejchanovaacute The physicochemical forms of tracers of radium in aqueous solutions containing chlorides sulfates and carbonates Radiochem Radioanal Letters 50(4) pp 227-242 1982

CREW 2006 ldquoComments on Radiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Technical Note No 2006-01 January 6 2006

Demir I Temporal variations and sources of Pb Co Cu Ni Fe and Mo in shallow ground water of the McDonalds Branch basin Lebanon State Forest Burlington County New Jersey New Jersey Geological Survey Report 20 19 p 1989

EPA 2000 Guidance for Data Quality Assessment Practical Methods for Data Analysis EPA QAG-9 July 2000

EPA 2002 Guidance for Comparing Background and Chemical Concentrations in Soil for CERCLA Site EPA 540-R-01-003 September 2002

FSP 2004 Field Sampling Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Gilbert 1989 Statistical Methods for Environmental Pollution Monitoring Richard O Gilbert 1987

Kirby H W M L Salutsky The Radiochemistry of Radium National Academy of Sciences ndash National Research Council NAS-NS 3057 205 p 1964

14

de maximis inc

Kramer T F and D F Reid The occurrence and behavior of radium in saline formation water of the US Gulf Coast Region Isotope Geoscience 2 pp 153shy174 1984

Landreacute V Comments on ldquoRadiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Nuclear Metals Superfund Site CREW Technical Note 2006-01 January 6 2006

Lauria D C and J M O Godoy Abnormal high natural radium concentration in surface waters Journal of Environmental Radioactivity 61(2) pp 27-37 2002

Lauria D C R M R Almeida O Sracek Behavior of radium thorium and uranium in groundwater near the Buena Lagoon in the Coastal Zone of the State of Rio de Janeiro Brazil Environmental Geology 47 pp 11-19 200

MACTEC 2005 Radiological Review of Fall 2004 RI Analytical Data Nuclear Metals Superfund Site Concord Massachusetts September 26 2004

MampE 2005 Memorandum from Andrew Schkuta to Ed Conroy Subject Uranium Analysis comparison of alpha spectrometry and ICPMS results December 15 2005

PSOP Project Summary and Operations Plan Nuclear Metals Superfund Site Concord Massachusetts April 15 2005

QAPP 2004 Quality Assurance Project Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Szabo Z D E Rice C L MacLeod and T H Barringer Relation of distribution of radium nitrate and pesticides to agricultural land use and depth Kirkwood-Cohansey aquifer system New Jersey Coastal Plain 1990-1991 US Geological Survey Water-Resources Investigations Report 96-4165A 119 p 1997

15

bull

bull bull

bull bull

bull bull bull

bull bull

bull bull

bull

500

1500

Figure 1

Con

cent

ratio

n (m

gkg

)

ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

1000

0

0 5 10 15 20 25 30 35 40

Samples

Figure 2

Alp

ha S

pect

osco

py (m

gkg

)

Fall 2004 Total Uranium No Super-Depleted

1000

100

10

1

y = 10604x R2 = 09049

1 10 100 1000

ICPMS (mgkg)

Fall 2004 Super-Depleted Linear Regression

Attachment 1Connolly Jayme

From Taylor Mel issagepam ail epa gov Sent Wednesday December 21 2005 932AM To brucetdemaximiscom jhuntdemaximiscom Subject Fw Uranium Analysis Issues

Rad-versus-ICPMS RADvsiCPMS_soilse MEvsPRP_ICPMSP memodoc (158 dPDF (29 KB) OF (25 KB)

brucejohn here is mampes take on the rns vs alpha spec analysis discrepancy

Forwarded by MelissaG TaylorR1USEPAUS on 12212005 0930 AM

Conroy Ed ltEdConroyregrn-ea ecomcomgt To

MelissaG TaylorR1USEPAUSregEPA 12152005 0416 cc PM

Subject FW Uranium Analysis Issues

Melissa

Heres Andys take on the alpha spec vs ICPMS debate along with a couple of graphs showing correlation between the two methods and MampE split data vs PRP data

Please call me on the cell phone tomorrow to discuss

Ed

Edward A Conroy PE

Project Manager

METCALF amp EDDYJAECOM

7812246880 ph

6172832942 cell

1

ed conroym-e aecom com Attachment 1

-----Original Message----shyFrom Schkuta Andrew Sent Thursday December 15 2005 358 PM To Conroy Ed Subject Uranium Analysis Issues

Ed

Please find attached a memorandum that summarizes my review of why ICP-MS and alpha spec data may not necessarily agree

ltltRad-versus-ICPMS memodocgtgt

In addition I have attached 2 preliminary charts that have the following draft information

ltltRADvsiCPMS soilsedPDFgtgt

The RADvsiCPMS soilsed chart is a draft plot of MampEs alpha spec versus ICP-MS results when data was available for both samples It shows the same general trend as the PRP data with fairly good agreement between the two analyses There appears to be a slight positive bias for alpha spec data at low concentrations but this is expected as discussed in my memo

ltltMEvsPRP ICPMSPDFgtgt

The MEvsPRP ICPMS is a draft plot of MampE vs the PRPs total uranium results as measured by ICP-MS -With the exception of a few outliers which will be further investigated there appears to be good agreement between MampEs and the PRPs data As a note the median percent difference between MampEs and the PRPs data was 1

Andrew Schkuta CHMM

Technical Specialist - Chemistry

METCALF amp EDDYJAECOM

701 Edgewater Drive

Wakefield MA 01880

(781) 224-6353

(781) 245-6293 fax

2

(See attached file Rad-versus-ICPMS Attachment 1memodoc) (See attached file RADvsiCPMS soilsedPDF) (See attached file MEvsPRP ICPMSPDF)

3

Attachment 1

Metcalf amp Eddy Inc

METCALFampEDDY

Memorandum

Date December 15 2005

To Ed Conroy

From Andrew Schkuta

Subject Uranium Analysis comparison of alpha spectrometry and ICP-MS results

File WA151 Nuclear Metals Distribution

Summary This memorandum discusses the differences between uranium analysis conducted by two different methods alpha spectroscopy (alpha spec) and inductively coupled plasma shymass spectrometry (ICP-MS) In summary the alpha spec data is expected to be biased high compared to the ICP-MS data but this is not indicative that either method is being performed incorrectly

In the September 25 2005 technical memorandum Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts (MACTEC 2005) data was presented comparing alpha spec and ICP-MS analysis for uranium analysis and it was proposed that future uranium analysis be performed by ICP-MS only For samples analyzed that were collected through Spring 2005 all samples were analyzed by ICP-MS with a subset (20) analyzed by alpha spec It was noted that there appeared to be a positive bias in the alpha spec data compared to the ICP-MS data and some concern has been voiced if analysis by ICP-MS only would meet project objectives This memorandum provides background on why the two analytical methods may yield differing results and discusses how they may impact project quality objectives

There are two primary reasons why alpha spec is expected to yield results that may be positively biased

1 Sample Preparation The sample preparation method for alpha spec is more rigorous The preparation method in alpha spec breaks down the silica mineral structure of the sample By design the ICP-MS sample preparation does not break down the mineral structure and is intended to remove all metals that could become environmentally available

2 Sample Quanitation The results for alpha spec are adjusted for analyte loss by correcting the result based on the amount of tracer recovered (yield) ICP-MS data is absolute ie data is not corrected for loss during sample preparation

2 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

These two reasons are discussed in further detail as follows

Sample Preparation

Radionuclide analysis traditionally uses a sample preparation method that includes hydrofluoric acid (HF) to break down the sample so that all of the uranium present in the sample is put into solution On the other hand standard EPA methods for metals analysis including ICP-MS do not measure the total metals present in a sample For example method SW846-3050 (EPA 1996) (as specified by the PRP in their approved QAPP for metals analysis by ICP-MS) states

This method is not a total digestion technique for most samples It is a very strong acid digestion that will dissolve almost all elements that could become environmentally available By design elements bound in silicate structures are not normally dissolved by this procedure as they are not usually mobile in the environment If absolute total digestion is required use Method 3052

As noted by the method a SW846 method is available to do a total digestion (equivalent to the alpha spec method) as method 3052 However method 3052 (EPA 1996) states (emphasis added)

Note This technique is not appropriate for regulatory applications that require the use of leachate preparations (ie Method 3050 Method 3051 Method 1311 Method 1312 Method 1310 Method 1320 Method 1330 Method 3031 Method 3040) This method is appropriate for those applications requiring a total decomposition for research purposes (ie~ geological studiesr mass balancesr analysis of Standard Reference Materials) or in response to a regulation that requires total sample decomposition

Even though the ICP-MS preparation method does not yield total results it still may conservatively measure the amount of metals that are environmentally available In the Research Project Summary Ambient Levels of Metals in New Jersey Soils prepared by the New Jersey Department of Environmental Protection (NJDEP 2003) where samples were prepared using SW846-3050 procedures it states

Metal concentrations reported in this summary were acid-extractable metals not necessarily total metal concentrations The extraction method (USEPA Method 3050) is a vigorous extraction method designed to remove all metals that could become environmentally available In practice the extraction method likely overestimates metals that could become available since the sample is refluxed with both concentrated nitric acid and hydrogen peroxide However by design the method will not extract chemicals incorporated in silica minerals as they are usually not mobile in the environment Thus concentrations reported in these studies may be lower than those indicated from analyses using methods designed to measure total metal concentrations such as x-ray fluorescence methods

Although references to uranium were not located it is expected that uranium will have some of the same type of mobility characteristics as other metals in the soil matrix The paper Ground Water Issue Behavior of Metals in Soil published by EPAs Superfund Technology Support Center for Ground Water (EPA 1992) describes how metals are distributed within soil The paper describes how metals are partitioned in various pools and how a rigorous extraction using HF is required to release metals present in the structure of primary or secondary minerals It is

3 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

expected that native metals (ie background concentrations) can be present in these structures The paper indicates that for four metals examined (cadmium copper zinc and nickel) the amount of metals that was bound in the mineral structure ranges from approximately 15-60 It would be expected that naturally occurring uranium would have similar characteristics

The National Institutes of Standards amp Technology (NIST) recognizes the differences between total digestion and method SW846-3050 and has issued Addendum to Certificates for the Certificate of Analysis (NIST 2003) for some Standard Reference Materials (SRMs) The Addendum to Certificates for SRM 2711 a Montana agricultural soil indicates that trace metals were measured at approximately 40 to 90 of their certified value when analyzed using SW846-3050 as the preparation method Even though uranium is not listed as a target analyte it is reasonable to expect that is would behave in a similar fashion

Sample Quanitation

The preparation method for alpha spec analysis includes addition of a tracer (U-232) to sample and is carried through the entire analytical scheme The amount of tracer is measured during analysis and the results for target compounds (the other uranium isotopes) are corrected for the amount of tracer recovery (yield) For example if amount of tracer measured is 50 of what was initially added to the sample the sample results are all corrected by dividing by 050 In terms of method quality assurancequality control (QAQC) the PRP QAPP indicates that tracer recovery in the range of 15-125 is considered acceptable and the analytical system is considered in control In other words up to 85 of the compound can be lost during sample preparation and analysis and the data is considered meeting all QAQC objectives This method of analysis is the equivalent to the technique of isotope dilution which is used for some non-radiological environmental analyses (eg EPA Method 1625 Semivolatile Organic Compounds by Isotope Dilution GCMS) Analyses using this technique although considered the most accurate are not widely employed due to the complexity and expense of using the isotope dilution technique

ICP-MS analysis measures the concentration of analyte with no correction for potential losses However QAQC procedures are in place to make sure the losses are not considered significant In addition to laboratory control samples matrix spike analysis measures the recovery for the target analyte that was over-spiked into the sample The PRP QAPP states that matrix spike recovery should be within the range of 75-125 Therefore up to 25 of the analyte could be lost and the method has still met the QAQC objectives If recovery is less than this then validation procedures acknowledge that the data may be biased low and sample results are flagged as estimated

Therefore because the alpha spec data is corrected for analyte loss during the analytical process while ICP-MS data is not some amount of positive bias is expected to be present in alpha spec data when compared to ICP-MS data

Considerations

In evaluating which analysis is appropriate for samples from the project neither method is inherently better than the other The following factors should be considered

1 What is being measured ICP-MS measures mass directly and data can be mathematically converted to activity Alpha spec measures activity directly and can be mathematically converted to mass

4 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

2 What are the objectives of the data in terms of measuring total versus environmentally available The ICP-MS method specified in the PRP QAPP measures uranium concentrations in the same manner that EPA has measured other metals concentrations Should uranium mass be measured in a typical way that will be used for measuring other metals of concern at the site (eg copper beryllium arsenic etc)

3 If ICP-MS is more cost effective that alpha spec should the option of using preparation method SW846-3052 be considered This should yield total uranium data similar to alpha spec analysis while still using the efficiencies of ICP-MS analysis

4 What units will the cleanup level for the site be in If its in mass units then ICPshyMS analysis may be more appropriate If its in activity units then alpha spec analysis may be warranted

References

Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts Technical Memorandum from Heath Downey to Bruce Thompson de maxim is inc September 26 2005 (MACTEC 2005)

Test Methods for Evaluating Solid Waste PhysicalChemical Methods SW-846 Third Edition with updates EPA 1996

Ambient Levels of Metals in New Jersey Soils Environmental Assessment and Risk Analysis Element Research Project Summary Division of Science Research and Technology State of New Jersey Department of Environmental Protection May2003

Ground Water Issue Behavior of Metals in Soil EPA Superfund Technology Support Center for Ground Water EPA540S-92018 October 1993

Certificate of Analysis Standard Reference Material 2711 Montana Soil National Institute of Standards amp Technology July 18 2003

bull bull

bull bull bullbull bull

10000

MampE SoilSediment Results Total Uranium mgKg

bull

bull - ~-~-----~~---~~-~--~~1000

en aE

I I bull Total a u

I -

bull bull 100 1shy -middot- -middot bullbull bull

Ybull bull

10 100 1000 10000

Alpha Spec

Attachm

ent 1

10 -

100000

TOTAL URANIUM BY ICP-MS mgKg

bull

bull 10000

I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 11: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

de maximis inc

It is also important to note that the historical data provided in the RI (PSOP 2005) was used to assist in design of the RI (FSP 2004) Data were compiled from many sources of information and represent samples collected analyzed and reported using various sample collection procedures laboratory analytical methods and data reporting conventions Data compilation activities involved both manual hand-entries and electronic entries when applicable Based on a review of the electronic database severe data quality limitations were noted including unknown analytical quality and validation levels missing location coordinates inconsistent media codes unknown detection limits and radionuclide uncertainty measurements inconsistent andor incorrect reporting of radionuclides using mass-based and activity-based units These limitations were considered significant during the scoping of the RIFS Work Plan therefore the historical data were used primarily to evaluate the spatial distribution of parameters and magnitude of detected concentrations The limitations of the historical data were used to provide the basis for identifying data gaps and scoping the RIFS As such the FSP included analysis for HTDRs at the areas of the Site known to be impacted by historical practices

With respect to comments provided on U-232 and U-236 these uranium isotopes are produced during the fission of uranium (as are plutonium americium and the rest of the HTDRs) Since none of these other fission products have been detected at the Site it is not credible that either uranium-232 or uranium-236 would be present

Response to Section 6 - Radium

The historical radium in groundwater data is suspect for several reasons including the previous discussion on historical data provided in the RI (PSOP 2005) Groundwater samples were collected from 13 wells at the Site (single sampling event) using bailers as part of this investigation All samples were screened for radioactivity by gross alpha beta analysis Only one sample exceeded the action level (gt 10 pCil gross alpha) with results of 50 pCil gross alpha and 55 pCil gross beta and it was analyzed for isotopic uranium Ra-226 and Ra-228 The results were 74 pCil U-234 025 pCil U-235 037 pCil U-238 12 pCil Ra-226 and 7 pCil Ra-228 These values are from Table 10 of the report and it appears that the Ra-226 value was incorrectly transcribed in the CREW Technical Note as 37 pCil The isotopic uranium results do not match those expected as U-234 is elevated and results for U-235 and U-238 are almost equal Therefore the analytical results for this sample are questionable Furthermore all the rest of the ground water samples had gross alpha results of 3 pCil or less indicating that there are no significant amounts of radium in groundwater

In addition a limited literature review was performed to assess the potential for leaching of radium from subsurface solids by nitrate at the NMI Site The literature demonstrates that radium may be found in a handful of aqueous phase complexes The predominant species of radium in aqueous solution is the divalent cation Ra2+ but other complexes such as RaOH+ RaCl+ RaCO3 and RaSO4 may also be found (Beneš et al 1982 Beneš 1984) Radium appears to be the least likely of all the alkali earth metals (beryllium magnesium calcium strontium barium and radium) to form aqueous phase

11

de maximis inc

complexes (Kirby and Salutsky 1964) As such we are unclear as to the origin of Dr Landreacutersquos statement that ldquonitrates are known to influence greatly radium leaching from the soil into the groundwaterrdquo (Landreacute 2006) The salt Ra(NO3)2 is highly soluble (245 g100 g H2O Kirby and Salutsky 1964) but we have not found evidence to suggest that the species Ra(NO3)+ or Ra(NO3)2 exist in aqueous solution Lauria et al (2004) present radionuclide and water chemistry data from groundwater samples taken from the Buena Lagoon in Brazil A plot of Ra-226 and Ra-228 concentrations versus nitrate concentration is presented in Figure 1 These data demonstrate that both Ra-226 and Ra-228 are almost completely uncorrelated with nitrate at this site as evidenced by correlation coefficients (r2) of approximately 005 and 004 respectively Szabo et al (1997) also report that no correlation exists between nitrate and Ra-226 in 17 groundwater samples taken from the Kirkwood-Cohansey aquifer system in New Jersey although Ra-228 appears to be weakly correlated with nitrate in the same data set (r2 = 018)

Figure 1 Correlation between aqueous radium and nitrate (Lauria et al 2004)

r 2 = 005

r 2 = 004

1E-10 1E-09 1E-08 1E-07 1E-06 1E-05 1E-04

0 5 10 15 20 25

Nitrate (mgL)

Ra

(microg

L)

Ra-226 Ra-228 Linear (Ra-226) Linear (Ra-228)

The mobility of radium in the subsurface appears to be controlled primarily by sorption and precipitation (Lauria et al 2004) High values of aqueous phase radium have been observed in groundwater samples with either increased salinity or reduced pH (Kraemer and Reed 1984 Lauria and Godoy 2002) Demir (1989) demonstrated the increased mobility of radium in acidic environments due to the reduced sorptive capacity of sediments at low pH Figure 2 is a plot of radium concentration versus sample pH from the same data set (Lauria et al 2004)

12

de maximis inc

Figure 2 Correlation between aqueous radium and pH (Lauria et al 2004) 226Ra = -4E-06pH + 3E-05

r2 = 058

228Ra = -5E-08pH + 4E-07 r2 = 032

1E-10 1E-09 1E-08 1E-07 1E-06 1E-05 1E-04

3 4 5 6 7

pH

Ra

(microg

L)

Ra-226 Ra-228 Linear (Ra-226) Linear (Ra-228)

Here a stronger correlation between Ra-226 concentration and pH exists (r2 = 058) along with an observable trend of decreasing aqueous phase radium with increasing pH Szabo et al (1997) also report a negative correlation (r2 = 030) between Ra-228 and pH but no significant correlation between Ra-226 and pH Divalent cations such as calcium and magnesium compete with radium for a limited number of surface complexation and adsorption sites (Szabo et al 1997) As lime (a mixture potentially containing calcium oxide calcium hydroxide and calcium carbonate) was co-disposed with uranium and nitric acid at the Holding Basin nitrate may serve as a site-specific tracer for conditions that may lead to the release of adsorbed or surface-complexed radium from the solid phase although no comparable calcium ldquoplumerdquo has been defined unlike nitrate in groundwater Therefore we suggest that Dr Landreacutersquos statement could be qualified by stating that low pH values are known to influence the leaching of radium from aquifer solids However as discussed above we do not see a compelling reason to add radium to the analytical list at this time

Please contact me if you have any questions

Sincerely

Bruce Thompson

cc Respondents Bennie Underwood PE de maximis inc inc Nelson Walter PE MACTEC Inc Peter Zeeb PhD PG GeoSyntec Consultants Inc

13

de maximis inc

Figure 1 ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

Figure 2 Comparison of Alpha Spectroscopy and ICP-MS Uranium Analysis

Attachment 1 Metcalf amp Eddy Memorandum ndash Uranium Analysis comparison of alpha spectrometry and ICP-MS results

Attachment 2 QAPP SOPs for Alpha Spectroscopy and ICPMS Sample Preparation

Attachment 3 Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor Department of the Army January 19 2000

References

Beneš P Migration of radium in the terrestrial hydrosphere The behavior of radium in waterways and aquifers Technical Reports Series No 310 International Atomic Energy Agency IAEA-TECDOC-301 Vienna Austrai pp 119-173 1984

Beneš P M Obdržaacutelek M Čejchanovaacute The physicochemical forms of tracers of radium in aqueous solutions containing chlorides sulfates and carbonates Radiochem Radioanal Letters 50(4) pp 227-242 1982

CREW 2006 ldquoComments on Radiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Technical Note No 2006-01 January 6 2006

Demir I Temporal variations and sources of Pb Co Cu Ni Fe and Mo in shallow ground water of the McDonalds Branch basin Lebanon State Forest Burlington County New Jersey New Jersey Geological Survey Report 20 19 p 1989

EPA 2000 Guidance for Data Quality Assessment Practical Methods for Data Analysis EPA QAG-9 July 2000

EPA 2002 Guidance for Comparing Background and Chemical Concentrations in Soil for CERCLA Site EPA 540-R-01-003 September 2002

FSP 2004 Field Sampling Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Gilbert 1989 Statistical Methods for Environmental Pollution Monitoring Richard O Gilbert 1987

Kirby H W M L Salutsky The Radiochemistry of Radium National Academy of Sciences ndash National Research Council NAS-NS 3057 205 p 1964

14

de maximis inc

Kramer T F and D F Reid The occurrence and behavior of radium in saline formation water of the US Gulf Coast Region Isotope Geoscience 2 pp 153shy174 1984

Landreacute V Comments on ldquoRadiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Nuclear Metals Superfund Site CREW Technical Note 2006-01 January 6 2006

Lauria D C and J M O Godoy Abnormal high natural radium concentration in surface waters Journal of Environmental Radioactivity 61(2) pp 27-37 2002

Lauria D C R M R Almeida O Sracek Behavior of radium thorium and uranium in groundwater near the Buena Lagoon in the Coastal Zone of the State of Rio de Janeiro Brazil Environmental Geology 47 pp 11-19 200

MACTEC 2005 Radiological Review of Fall 2004 RI Analytical Data Nuclear Metals Superfund Site Concord Massachusetts September 26 2004

MampE 2005 Memorandum from Andrew Schkuta to Ed Conroy Subject Uranium Analysis comparison of alpha spectrometry and ICPMS results December 15 2005

PSOP Project Summary and Operations Plan Nuclear Metals Superfund Site Concord Massachusetts April 15 2005

QAPP 2004 Quality Assurance Project Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Szabo Z D E Rice C L MacLeod and T H Barringer Relation of distribution of radium nitrate and pesticides to agricultural land use and depth Kirkwood-Cohansey aquifer system New Jersey Coastal Plain 1990-1991 US Geological Survey Water-Resources Investigations Report 96-4165A 119 p 1997

15

bull

bull bull

bull bull

bull bull bull

bull bull

bull bull

bull

500

1500

Figure 1

Con

cent

ratio

n (m

gkg

)

ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

1000

0

0 5 10 15 20 25 30 35 40

Samples

Figure 2

Alp

ha S

pect

osco

py (m

gkg

)

Fall 2004 Total Uranium No Super-Depleted

1000

100

10

1

y = 10604x R2 = 09049

1 10 100 1000

ICPMS (mgkg)

Fall 2004 Super-Depleted Linear Regression

Attachment 1Connolly Jayme

From Taylor Mel issagepam ail epa gov Sent Wednesday December 21 2005 932AM To brucetdemaximiscom jhuntdemaximiscom Subject Fw Uranium Analysis Issues

Rad-versus-ICPMS RADvsiCPMS_soilse MEvsPRP_ICPMSP memodoc (158 dPDF (29 KB) OF (25 KB)

brucejohn here is mampes take on the rns vs alpha spec analysis discrepancy

Forwarded by MelissaG TaylorR1USEPAUS on 12212005 0930 AM

Conroy Ed ltEdConroyregrn-ea ecomcomgt To

MelissaG TaylorR1USEPAUSregEPA 12152005 0416 cc PM

Subject FW Uranium Analysis Issues

Melissa

Heres Andys take on the alpha spec vs ICPMS debate along with a couple of graphs showing correlation between the two methods and MampE split data vs PRP data

Please call me on the cell phone tomorrow to discuss

Ed

Edward A Conroy PE

Project Manager

METCALF amp EDDYJAECOM

7812246880 ph

6172832942 cell

1

ed conroym-e aecom com Attachment 1

-----Original Message----shyFrom Schkuta Andrew Sent Thursday December 15 2005 358 PM To Conroy Ed Subject Uranium Analysis Issues

Ed

Please find attached a memorandum that summarizes my review of why ICP-MS and alpha spec data may not necessarily agree

ltltRad-versus-ICPMS memodocgtgt

In addition I have attached 2 preliminary charts that have the following draft information

ltltRADvsiCPMS soilsedPDFgtgt

The RADvsiCPMS soilsed chart is a draft plot of MampEs alpha spec versus ICP-MS results when data was available for both samples It shows the same general trend as the PRP data with fairly good agreement between the two analyses There appears to be a slight positive bias for alpha spec data at low concentrations but this is expected as discussed in my memo

ltltMEvsPRP ICPMSPDFgtgt

The MEvsPRP ICPMS is a draft plot of MampE vs the PRPs total uranium results as measured by ICP-MS -With the exception of a few outliers which will be further investigated there appears to be good agreement between MampEs and the PRPs data As a note the median percent difference between MampEs and the PRPs data was 1

Andrew Schkuta CHMM

Technical Specialist - Chemistry

METCALF amp EDDYJAECOM

701 Edgewater Drive

Wakefield MA 01880

(781) 224-6353

(781) 245-6293 fax

2

(See attached file Rad-versus-ICPMS Attachment 1memodoc) (See attached file RADvsiCPMS soilsedPDF) (See attached file MEvsPRP ICPMSPDF)

3

Attachment 1

Metcalf amp Eddy Inc

METCALFampEDDY

Memorandum

Date December 15 2005

To Ed Conroy

From Andrew Schkuta

Subject Uranium Analysis comparison of alpha spectrometry and ICP-MS results

File WA151 Nuclear Metals Distribution

Summary This memorandum discusses the differences between uranium analysis conducted by two different methods alpha spectroscopy (alpha spec) and inductively coupled plasma shymass spectrometry (ICP-MS) In summary the alpha spec data is expected to be biased high compared to the ICP-MS data but this is not indicative that either method is being performed incorrectly

In the September 25 2005 technical memorandum Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts (MACTEC 2005) data was presented comparing alpha spec and ICP-MS analysis for uranium analysis and it was proposed that future uranium analysis be performed by ICP-MS only For samples analyzed that were collected through Spring 2005 all samples were analyzed by ICP-MS with a subset (20) analyzed by alpha spec It was noted that there appeared to be a positive bias in the alpha spec data compared to the ICP-MS data and some concern has been voiced if analysis by ICP-MS only would meet project objectives This memorandum provides background on why the two analytical methods may yield differing results and discusses how they may impact project quality objectives

There are two primary reasons why alpha spec is expected to yield results that may be positively biased

1 Sample Preparation The sample preparation method for alpha spec is more rigorous The preparation method in alpha spec breaks down the silica mineral structure of the sample By design the ICP-MS sample preparation does not break down the mineral structure and is intended to remove all metals that could become environmentally available

2 Sample Quanitation The results for alpha spec are adjusted for analyte loss by correcting the result based on the amount of tracer recovered (yield) ICP-MS data is absolute ie data is not corrected for loss during sample preparation

2 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

These two reasons are discussed in further detail as follows

Sample Preparation

Radionuclide analysis traditionally uses a sample preparation method that includes hydrofluoric acid (HF) to break down the sample so that all of the uranium present in the sample is put into solution On the other hand standard EPA methods for metals analysis including ICP-MS do not measure the total metals present in a sample For example method SW846-3050 (EPA 1996) (as specified by the PRP in their approved QAPP for metals analysis by ICP-MS) states

This method is not a total digestion technique for most samples It is a very strong acid digestion that will dissolve almost all elements that could become environmentally available By design elements bound in silicate structures are not normally dissolved by this procedure as they are not usually mobile in the environment If absolute total digestion is required use Method 3052

As noted by the method a SW846 method is available to do a total digestion (equivalent to the alpha spec method) as method 3052 However method 3052 (EPA 1996) states (emphasis added)

Note This technique is not appropriate for regulatory applications that require the use of leachate preparations (ie Method 3050 Method 3051 Method 1311 Method 1312 Method 1310 Method 1320 Method 1330 Method 3031 Method 3040) This method is appropriate for those applications requiring a total decomposition for research purposes (ie~ geological studiesr mass balancesr analysis of Standard Reference Materials) or in response to a regulation that requires total sample decomposition

Even though the ICP-MS preparation method does not yield total results it still may conservatively measure the amount of metals that are environmentally available In the Research Project Summary Ambient Levels of Metals in New Jersey Soils prepared by the New Jersey Department of Environmental Protection (NJDEP 2003) where samples were prepared using SW846-3050 procedures it states

Metal concentrations reported in this summary were acid-extractable metals not necessarily total metal concentrations The extraction method (USEPA Method 3050) is a vigorous extraction method designed to remove all metals that could become environmentally available In practice the extraction method likely overestimates metals that could become available since the sample is refluxed with both concentrated nitric acid and hydrogen peroxide However by design the method will not extract chemicals incorporated in silica minerals as they are usually not mobile in the environment Thus concentrations reported in these studies may be lower than those indicated from analyses using methods designed to measure total metal concentrations such as x-ray fluorescence methods

Although references to uranium were not located it is expected that uranium will have some of the same type of mobility characteristics as other metals in the soil matrix The paper Ground Water Issue Behavior of Metals in Soil published by EPAs Superfund Technology Support Center for Ground Water (EPA 1992) describes how metals are distributed within soil The paper describes how metals are partitioned in various pools and how a rigorous extraction using HF is required to release metals present in the structure of primary or secondary minerals It is

3 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

expected that native metals (ie background concentrations) can be present in these structures The paper indicates that for four metals examined (cadmium copper zinc and nickel) the amount of metals that was bound in the mineral structure ranges from approximately 15-60 It would be expected that naturally occurring uranium would have similar characteristics

The National Institutes of Standards amp Technology (NIST) recognizes the differences between total digestion and method SW846-3050 and has issued Addendum to Certificates for the Certificate of Analysis (NIST 2003) for some Standard Reference Materials (SRMs) The Addendum to Certificates for SRM 2711 a Montana agricultural soil indicates that trace metals were measured at approximately 40 to 90 of their certified value when analyzed using SW846-3050 as the preparation method Even though uranium is not listed as a target analyte it is reasonable to expect that is would behave in a similar fashion

Sample Quanitation

The preparation method for alpha spec analysis includes addition of a tracer (U-232) to sample and is carried through the entire analytical scheme The amount of tracer is measured during analysis and the results for target compounds (the other uranium isotopes) are corrected for the amount of tracer recovery (yield) For example if amount of tracer measured is 50 of what was initially added to the sample the sample results are all corrected by dividing by 050 In terms of method quality assurancequality control (QAQC) the PRP QAPP indicates that tracer recovery in the range of 15-125 is considered acceptable and the analytical system is considered in control In other words up to 85 of the compound can be lost during sample preparation and analysis and the data is considered meeting all QAQC objectives This method of analysis is the equivalent to the technique of isotope dilution which is used for some non-radiological environmental analyses (eg EPA Method 1625 Semivolatile Organic Compounds by Isotope Dilution GCMS) Analyses using this technique although considered the most accurate are not widely employed due to the complexity and expense of using the isotope dilution technique

ICP-MS analysis measures the concentration of analyte with no correction for potential losses However QAQC procedures are in place to make sure the losses are not considered significant In addition to laboratory control samples matrix spike analysis measures the recovery for the target analyte that was over-spiked into the sample The PRP QAPP states that matrix spike recovery should be within the range of 75-125 Therefore up to 25 of the analyte could be lost and the method has still met the QAQC objectives If recovery is less than this then validation procedures acknowledge that the data may be biased low and sample results are flagged as estimated

Therefore because the alpha spec data is corrected for analyte loss during the analytical process while ICP-MS data is not some amount of positive bias is expected to be present in alpha spec data when compared to ICP-MS data

Considerations

In evaluating which analysis is appropriate for samples from the project neither method is inherently better than the other The following factors should be considered

1 What is being measured ICP-MS measures mass directly and data can be mathematically converted to activity Alpha spec measures activity directly and can be mathematically converted to mass

4 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

2 What are the objectives of the data in terms of measuring total versus environmentally available The ICP-MS method specified in the PRP QAPP measures uranium concentrations in the same manner that EPA has measured other metals concentrations Should uranium mass be measured in a typical way that will be used for measuring other metals of concern at the site (eg copper beryllium arsenic etc)

3 If ICP-MS is more cost effective that alpha spec should the option of using preparation method SW846-3052 be considered This should yield total uranium data similar to alpha spec analysis while still using the efficiencies of ICP-MS analysis

4 What units will the cleanup level for the site be in If its in mass units then ICPshyMS analysis may be more appropriate If its in activity units then alpha spec analysis may be warranted

References

Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts Technical Memorandum from Heath Downey to Bruce Thompson de maxim is inc September 26 2005 (MACTEC 2005)

Test Methods for Evaluating Solid Waste PhysicalChemical Methods SW-846 Third Edition with updates EPA 1996

Ambient Levels of Metals in New Jersey Soils Environmental Assessment and Risk Analysis Element Research Project Summary Division of Science Research and Technology State of New Jersey Department of Environmental Protection May2003

Ground Water Issue Behavior of Metals in Soil EPA Superfund Technology Support Center for Ground Water EPA540S-92018 October 1993

Certificate of Analysis Standard Reference Material 2711 Montana Soil National Institute of Standards amp Technology July 18 2003

bull bull

bull bull bullbull bull

10000

MampE SoilSediment Results Total Uranium mgKg

bull

bull - ~-~-----~~---~~-~--~~1000

en aE

I I bull Total a u

I -

bull bull 100 1shy -middot- -middot bullbull bull

Ybull bull

10 100 1000 10000

Alpha Spec

Attachm

ent 1

10 -

100000

TOTAL URANIUM BY ICP-MS mgKg

bull

bull 10000

I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 12: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

de maximis inc

complexes (Kirby and Salutsky 1964) As such we are unclear as to the origin of Dr Landreacutersquos statement that ldquonitrates are known to influence greatly radium leaching from the soil into the groundwaterrdquo (Landreacute 2006) The salt Ra(NO3)2 is highly soluble (245 g100 g H2O Kirby and Salutsky 1964) but we have not found evidence to suggest that the species Ra(NO3)+ or Ra(NO3)2 exist in aqueous solution Lauria et al (2004) present radionuclide and water chemistry data from groundwater samples taken from the Buena Lagoon in Brazil A plot of Ra-226 and Ra-228 concentrations versus nitrate concentration is presented in Figure 1 These data demonstrate that both Ra-226 and Ra-228 are almost completely uncorrelated with nitrate at this site as evidenced by correlation coefficients (r2) of approximately 005 and 004 respectively Szabo et al (1997) also report that no correlation exists between nitrate and Ra-226 in 17 groundwater samples taken from the Kirkwood-Cohansey aquifer system in New Jersey although Ra-228 appears to be weakly correlated with nitrate in the same data set (r2 = 018)

Figure 1 Correlation between aqueous radium and nitrate (Lauria et al 2004)

r 2 = 005

r 2 = 004

1E-10 1E-09 1E-08 1E-07 1E-06 1E-05 1E-04

0 5 10 15 20 25

Nitrate (mgL)

Ra

(microg

L)

Ra-226 Ra-228 Linear (Ra-226) Linear (Ra-228)

The mobility of radium in the subsurface appears to be controlled primarily by sorption and precipitation (Lauria et al 2004) High values of aqueous phase radium have been observed in groundwater samples with either increased salinity or reduced pH (Kraemer and Reed 1984 Lauria and Godoy 2002) Demir (1989) demonstrated the increased mobility of radium in acidic environments due to the reduced sorptive capacity of sediments at low pH Figure 2 is a plot of radium concentration versus sample pH from the same data set (Lauria et al 2004)

12

de maximis inc

Figure 2 Correlation between aqueous radium and pH (Lauria et al 2004) 226Ra = -4E-06pH + 3E-05

r2 = 058

228Ra = -5E-08pH + 4E-07 r2 = 032

1E-10 1E-09 1E-08 1E-07 1E-06 1E-05 1E-04

3 4 5 6 7

pH

Ra

(microg

L)

Ra-226 Ra-228 Linear (Ra-226) Linear (Ra-228)

Here a stronger correlation between Ra-226 concentration and pH exists (r2 = 058) along with an observable trend of decreasing aqueous phase radium with increasing pH Szabo et al (1997) also report a negative correlation (r2 = 030) between Ra-228 and pH but no significant correlation between Ra-226 and pH Divalent cations such as calcium and magnesium compete with radium for a limited number of surface complexation and adsorption sites (Szabo et al 1997) As lime (a mixture potentially containing calcium oxide calcium hydroxide and calcium carbonate) was co-disposed with uranium and nitric acid at the Holding Basin nitrate may serve as a site-specific tracer for conditions that may lead to the release of adsorbed or surface-complexed radium from the solid phase although no comparable calcium ldquoplumerdquo has been defined unlike nitrate in groundwater Therefore we suggest that Dr Landreacutersquos statement could be qualified by stating that low pH values are known to influence the leaching of radium from aquifer solids However as discussed above we do not see a compelling reason to add radium to the analytical list at this time

Please contact me if you have any questions

Sincerely

Bruce Thompson

cc Respondents Bennie Underwood PE de maximis inc inc Nelson Walter PE MACTEC Inc Peter Zeeb PhD PG GeoSyntec Consultants Inc

13

de maximis inc

Figure 1 ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

Figure 2 Comparison of Alpha Spectroscopy and ICP-MS Uranium Analysis

Attachment 1 Metcalf amp Eddy Memorandum ndash Uranium Analysis comparison of alpha spectrometry and ICP-MS results

Attachment 2 QAPP SOPs for Alpha Spectroscopy and ICPMS Sample Preparation

Attachment 3 Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor Department of the Army January 19 2000

References

Beneš P Migration of radium in the terrestrial hydrosphere The behavior of radium in waterways and aquifers Technical Reports Series No 310 International Atomic Energy Agency IAEA-TECDOC-301 Vienna Austrai pp 119-173 1984

Beneš P M Obdržaacutelek M Čejchanovaacute The physicochemical forms of tracers of radium in aqueous solutions containing chlorides sulfates and carbonates Radiochem Radioanal Letters 50(4) pp 227-242 1982

CREW 2006 ldquoComments on Radiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Technical Note No 2006-01 January 6 2006

Demir I Temporal variations and sources of Pb Co Cu Ni Fe and Mo in shallow ground water of the McDonalds Branch basin Lebanon State Forest Burlington County New Jersey New Jersey Geological Survey Report 20 19 p 1989

EPA 2000 Guidance for Data Quality Assessment Practical Methods for Data Analysis EPA QAG-9 July 2000

EPA 2002 Guidance for Comparing Background and Chemical Concentrations in Soil for CERCLA Site EPA 540-R-01-003 September 2002

FSP 2004 Field Sampling Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Gilbert 1989 Statistical Methods for Environmental Pollution Monitoring Richard O Gilbert 1987

Kirby H W M L Salutsky The Radiochemistry of Radium National Academy of Sciences ndash National Research Council NAS-NS 3057 205 p 1964

14

de maximis inc

Kramer T F and D F Reid The occurrence and behavior of radium in saline formation water of the US Gulf Coast Region Isotope Geoscience 2 pp 153shy174 1984

Landreacute V Comments on ldquoRadiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Nuclear Metals Superfund Site CREW Technical Note 2006-01 January 6 2006

Lauria D C and J M O Godoy Abnormal high natural radium concentration in surface waters Journal of Environmental Radioactivity 61(2) pp 27-37 2002

Lauria D C R M R Almeida O Sracek Behavior of radium thorium and uranium in groundwater near the Buena Lagoon in the Coastal Zone of the State of Rio de Janeiro Brazil Environmental Geology 47 pp 11-19 200

MACTEC 2005 Radiological Review of Fall 2004 RI Analytical Data Nuclear Metals Superfund Site Concord Massachusetts September 26 2004

MampE 2005 Memorandum from Andrew Schkuta to Ed Conroy Subject Uranium Analysis comparison of alpha spectrometry and ICPMS results December 15 2005

PSOP Project Summary and Operations Plan Nuclear Metals Superfund Site Concord Massachusetts April 15 2005

QAPP 2004 Quality Assurance Project Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Szabo Z D E Rice C L MacLeod and T H Barringer Relation of distribution of radium nitrate and pesticides to agricultural land use and depth Kirkwood-Cohansey aquifer system New Jersey Coastal Plain 1990-1991 US Geological Survey Water-Resources Investigations Report 96-4165A 119 p 1997

15

bull

bull bull

bull bull

bull bull bull

bull bull

bull bull

bull

500

1500

Figure 1

Con

cent

ratio

n (m

gkg

)

ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

1000

0

0 5 10 15 20 25 30 35 40

Samples

Figure 2

Alp

ha S

pect

osco

py (m

gkg

)

Fall 2004 Total Uranium No Super-Depleted

1000

100

10

1

y = 10604x R2 = 09049

1 10 100 1000

ICPMS (mgkg)

Fall 2004 Super-Depleted Linear Regression

Attachment 1Connolly Jayme

From Taylor Mel issagepam ail epa gov Sent Wednesday December 21 2005 932AM To brucetdemaximiscom jhuntdemaximiscom Subject Fw Uranium Analysis Issues

Rad-versus-ICPMS RADvsiCPMS_soilse MEvsPRP_ICPMSP memodoc (158 dPDF (29 KB) OF (25 KB)

brucejohn here is mampes take on the rns vs alpha spec analysis discrepancy

Forwarded by MelissaG TaylorR1USEPAUS on 12212005 0930 AM

Conroy Ed ltEdConroyregrn-ea ecomcomgt To

MelissaG TaylorR1USEPAUSregEPA 12152005 0416 cc PM

Subject FW Uranium Analysis Issues

Melissa

Heres Andys take on the alpha spec vs ICPMS debate along with a couple of graphs showing correlation between the two methods and MampE split data vs PRP data

Please call me on the cell phone tomorrow to discuss

Ed

Edward A Conroy PE

Project Manager

METCALF amp EDDYJAECOM

7812246880 ph

6172832942 cell

1

ed conroym-e aecom com Attachment 1

-----Original Message----shyFrom Schkuta Andrew Sent Thursday December 15 2005 358 PM To Conroy Ed Subject Uranium Analysis Issues

Ed

Please find attached a memorandum that summarizes my review of why ICP-MS and alpha spec data may not necessarily agree

ltltRad-versus-ICPMS memodocgtgt

In addition I have attached 2 preliminary charts that have the following draft information

ltltRADvsiCPMS soilsedPDFgtgt

The RADvsiCPMS soilsed chart is a draft plot of MampEs alpha spec versus ICP-MS results when data was available for both samples It shows the same general trend as the PRP data with fairly good agreement between the two analyses There appears to be a slight positive bias for alpha spec data at low concentrations but this is expected as discussed in my memo

ltltMEvsPRP ICPMSPDFgtgt

The MEvsPRP ICPMS is a draft plot of MampE vs the PRPs total uranium results as measured by ICP-MS -With the exception of a few outliers which will be further investigated there appears to be good agreement between MampEs and the PRPs data As a note the median percent difference between MampEs and the PRPs data was 1

Andrew Schkuta CHMM

Technical Specialist - Chemistry

METCALF amp EDDYJAECOM

701 Edgewater Drive

Wakefield MA 01880

(781) 224-6353

(781) 245-6293 fax

2

(See attached file Rad-versus-ICPMS Attachment 1memodoc) (See attached file RADvsiCPMS soilsedPDF) (See attached file MEvsPRP ICPMSPDF)

3

Attachment 1

Metcalf amp Eddy Inc

METCALFampEDDY

Memorandum

Date December 15 2005

To Ed Conroy

From Andrew Schkuta

Subject Uranium Analysis comparison of alpha spectrometry and ICP-MS results

File WA151 Nuclear Metals Distribution

Summary This memorandum discusses the differences between uranium analysis conducted by two different methods alpha spectroscopy (alpha spec) and inductively coupled plasma shymass spectrometry (ICP-MS) In summary the alpha spec data is expected to be biased high compared to the ICP-MS data but this is not indicative that either method is being performed incorrectly

In the September 25 2005 technical memorandum Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts (MACTEC 2005) data was presented comparing alpha spec and ICP-MS analysis for uranium analysis and it was proposed that future uranium analysis be performed by ICP-MS only For samples analyzed that were collected through Spring 2005 all samples were analyzed by ICP-MS with a subset (20) analyzed by alpha spec It was noted that there appeared to be a positive bias in the alpha spec data compared to the ICP-MS data and some concern has been voiced if analysis by ICP-MS only would meet project objectives This memorandum provides background on why the two analytical methods may yield differing results and discusses how they may impact project quality objectives

There are two primary reasons why alpha spec is expected to yield results that may be positively biased

1 Sample Preparation The sample preparation method for alpha spec is more rigorous The preparation method in alpha spec breaks down the silica mineral structure of the sample By design the ICP-MS sample preparation does not break down the mineral structure and is intended to remove all metals that could become environmentally available

2 Sample Quanitation The results for alpha spec are adjusted for analyte loss by correcting the result based on the amount of tracer recovered (yield) ICP-MS data is absolute ie data is not corrected for loss during sample preparation

2 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

These two reasons are discussed in further detail as follows

Sample Preparation

Radionuclide analysis traditionally uses a sample preparation method that includes hydrofluoric acid (HF) to break down the sample so that all of the uranium present in the sample is put into solution On the other hand standard EPA methods for metals analysis including ICP-MS do not measure the total metals present in a sample For example method SW846-3050 (EPA 1996) (as specified by the PRP in their approved QAPP for metals analysis by ICP-MS) states

This method is not a total digestion technique for most samples It is a very strong acid digestion that will dissolve almost all elements that could become environmentally available By design elements bound in silicate structures are not normally dissolved by this procedure as they are not usually mobile in the environment If absolute total digestion is required use Method 3052

As noted by the method a SW846 method is available to do a total digestion (equivalent to the alpha spec method) as method 3052 However method 3052 (EPA 1996) states (emphasis added)

Note This technique is not appropriate for regulatory applications that require the use of leachate preparations (ie Method 3050 Method 3051 Method 1311 Method 1312 Method 1310 Method 1320 Method 1330 Method 3031 Method 3040) This method is appropriate for those applications requiring a total decomposition for research purposes (ie~ geological studiesr mass balancesr analysis of Standard Reference Materials) or in response to a regulation that requires total sample decomposition

Even though the ICP-MS preparation method does not yield total results it still may conservatively measure the amount of metals that are environmentally available In the Research Project Summary Ambient Levels of Metals in New Jersey Soils prepared by the New Jersey Department of Environmental Protection (NJDEP 2003) where samples were prepared using SW846-3050 procedures it states

Metal concentrations reported in this summary were acid-extractable metals not necessarily total metal concentrations The extraction method (USEPA Method 3050) is a vigorous extraction method designed to remove all metals that could become environmentally available In practice the extraction method likely overestimates metals that could become available since the sample is refluxed with both concentrated nitric acid and hydrogen peroxide However by design the method will not extract chemicals incorporated in silica minerals as they are usually not mobile in the environment Thus concentrations reported in these studies may be lower than those indicated from analyses using methods designed to measure total metal concentrations such as x-ray fluorescence methods

Although references to uranium were not located it is expected that uranium will have some of the same type of mobility characteristics as other metals in the soil matrix The paper Ground Water Issue Behavior of Metals in Soil published by EPAs Superfund Technology Support Center for Ground Water (EPA 1992) describes how metals are distributed within soil The paper describes how metals are partitioned in various pools and how a rigorous extraction using HF is required to release metals present in the structure of primary or secondary minerals It is

3 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

expected that native metals (ie background concentrations) can be present in these structures The paper indicates that for four metals examined (cadmium copper zinc and nickel) the amount of metals that was bound in the mineral structure ranges from approximately 15-60 It would be expected that naturally occurring uranium would have similar characteristics

The National Institutes of Standards amp Technology (NIST) recognizes the differences between total digestion and method SW846-3050 and has issued Addendum to Certificates for the Certificate of Analysis (NIST 2003) for some Standard Reference Materials (SRMs) The Addendum to Certificates for SRM 2711 a Montana agricultural soil indicates that trace metals were measured at approximately 40 to 90 of their certified value when analyzed using SW846-3050 as the preparation method Even though uranium is not listed as a target analyte it is reasonable to expect that is would behave in a similar fashion

Sample Quanitation

The preparation method for alpha spec analysis includes addition of a tracer (U-232) to sample and is carried through the entire analytical scheme The amount of tracer is measured during analysis and the results for target compounds (the other uranium isotopes) are corrected for the amount of tracer recovery (yield) For example if amount of tracer measured is 50 of what was initially added to the sample the sample results are all corrected by dividing by 050 In terms of method quality assurancequality control (QAQC) the PRP QAPP indicates that tracer recovery in the range of 15-125 is considered acceptable and the analytical system is considered in control In other words up to 85 of the compound can be lost during sample preparation and analysis and the data is considered meeting all QAQC objectives This method of analysis is the equivalent to the technique of isotope dilution which is used for some non-radiological environmental analyses (eg EPA Method 1625 Semivolatile Organic Compounds by Isotope Dilution GCMS) Analyses using this technique although considered the most accurate are not widely employed due to the complexity and expense of using the isotope dilution technique

ICP-MS analysis measures the concentration of analyte with no correction for potential losses However QAQC procedures are in place to make sure the losses are not considered significant In addition to laboratory control samples matrix spike analysis measures the recovery for the target analyte that was over-spiked into the sample The PRP QAPP states that matrix spike recovery should be within the range of 75-125 Therefore up to 25 of the analyte could be lost and the method has still met the QAQC objectives If recovery is less than this then validation procedures acknowledge that the data may be biased low and sample results are flagged as estimated

Therefore because the alpha spec data is corrected for analyte loss during the analytical process while ICP-MS data is not some amount of positive bias is expected to be present in alpha spec data when compared to ICP-MS data

Considerations

In evaluating which analysis is appropriate for samples from the project neither method is inherently better than the other The following factors should be considered

1 What is being measured ICP-MS measures mass directly and data can be mathematically converted to activity Alpha spec measures activity directly and can be mathematically converted to mass

4 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

2 What are the objectives of the data in terms of measuring total versus environmentally available The ICP-MS method specified in the PRP QAPP measures uranium concentrations in the same manner that EPA has measured other metals concentrations Should uranium mass be measured in a typical way that will be used for measuring other metals of concern at the site (eg copper beryllium arsenic etc)

3 If ICP-MS is more cost effective that alpha spec should the option of using preparation method SW846-3052 be considered This should yield total uranium data similar to alpha spec analysis while still using the efficiencies of ICP-MS analysis

4 What units will the cleanup level for the site be in If its in mass units then ICPshyMS analysis may be more appropriate If its in activity units then alpha spec analysis may be warranted

References

Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts Technical Memorandum from Heath Downey to Bruce Thompson de maxim is inc September 26 2005 (MACTEC 2005)

Test Methods for Evaluating Solid Waste PhysicalChemical Methods SW-846 Third Edition with updates EPA 1996

Ambient Levels of Metals in New Jersey Soils Environmental Assessment and Risk Analysis Element Research Project Summary Division of Science Research and Technology State of New Jersey Department of Environmental Protection May2003

Ground Water Issue Behavior of Metals in Soil EPA Superfund Technology Support Center for Ground Water EPA540S-92018 October 1993

Certificate of Analysis Standard Reference Material 2711 Montana Soil National Institute of Standards amp Technology July 18 2003

bull bull

bull bull bullbull bull

10000

MampE SoilSediment Results Total Uranium mgKg

bull

bull - ~-~-----~~---~~-~--~~1000

en aE

I I bull Total a u

I -

bull bull 100 1shy -middot- -middot bullbull bull

Ybull bull

10 100 1000 10000

Alpha Spec

Attachm

ent 1

10 -

100000

TOTAL URANIUM BY ICP-MS mgKg

bull

bull 10000

I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 13: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

de maximis inc

Figure 2 Correlation between aqueous radium and pH (Lauria et al 2004) 226Ra = -4E-06pH + 3E-05

r2 = 058

228Ra = -5E-08pH + 4E-07 r2 = 032

1E-10 1E-09 1E-08 1E-07 1E-06 1E-05 1E-04

3 4 5 6 7

pH

Ra

(microg

L)

Ra-226 Ra-228 Linear (Ra-226) Linear (Ra-228)

Here a stronger correlation between Ra-226 concentration and pH exists (r2 = 058) along with an observable trend of decreasing aqueous phase radium with increasing pH Szabo et al (1997) also report a negative correlation (r2 = 030) between Ra-228 and pH but no significant correlation between Ra-226 and pH Divalent cations such as calcium and magnesium compete with radium for a limited number of surface complexation and adsorption sites (Szabo et al 1997) As lime (a mixture potentially containing calcium oxide calcium hydroxide and calcium carbonate) was co-disposed with uranium and nitric acid at the Holding Basin nitrate may serve as a site-specific tracer for conditions that may lead to the release of adsorbed or surface-complexed radium from the solid phase although no comparable calcium ldquoplumerdquo has been defined unlike nitrate in groundwater Therefore we suggest that Dr Landreacutersquos statement could be qualified by stating that low pH values are known to influence the leaching of radium from aquifer solids However as discussed above we do not see a compelling reason to add radium to the analytical list at this time

Please contact me if you have any questions

Sincerely

Bruce Thompson

cc Respondents Bennie Underwood PE de maximis inc inc Nelson Walter PE MACTEC Inc Peter Zeeb PhD PG GeoSyntec Consultants Inc

13

de maximis inc

Figure 1 ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

Figure 2 Comparison of Alpha Spectroscopy and ICP-MS Uranium Analysis

Attachment 1 Metcalf amp Eddy Memorandum ndash Uranium Analysis comparison of alpha spectrometry and ICP-MS results

Attachment 2 QAPP SOPs for Alpha Spectroscopy and ICPMS Sample Preparation

Attachment 3 Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor Department of the Army January 19 2000

References

Beneš P Migration of radium in the terrestrial hydrosphere The behavior of radium in waterways and aquifers Technical Reports Series No 310 International Atomic Energy Agency IAEA-TECDOC-301 Vienna Austrai pp 119-173 1984

Beneš P M Obdržaacutelek M Čejchanovaacute The physicochemical forms of tracers of radium in aqueous solutions containing chlorides sulfates and carbonates Radiochem Radioanal Letters 50(4) pp 227-242 1982

CREW 2006 ldquoComments on Radiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Technical Note No 2006-01 January 6 2006

Demir I Temporal variations and sources of Pb Co Cu Ni Fe and Mo in shallow ground water of the McDonalds Branch basin Lebanon State Forest Burlington County New Jersey New Jersey Geological Survey Report 20 19 p 1989

EPA 2000 Guidance for Data Quality Assessment Practical Methods for Data Analysis EPA QAG-9 July 2000

EPA 2002 Guidance for Comparing Background and Chemical Concentrations in Soil for CERCLA Site EPA 540-R-01-003 September 2002

FSP 2004 Field Sampling Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Gilbert 1989 Statistical Methods for Environmental Pollution Monitoring Richard O Gilbert 1987

Kirby H W M L Salutsky The Radiochemistry of Radium National Academy of Sciences ndash National Research Council NAS-NS 3057 205 p 1964

14

de maximis inc

Kramer T F and D F Reid The occurrence and behavior of radium in saline formation water of the US Gulf Coast Region Isotope Geoscience 2 pp 153shy174 1984

Landreacute V Comments on ldquoRadiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Nuclear Metals Superfund Site CREW Technical Note 2006-01 January 6 2006

Lauria D C and J M O Godoy Abnormal high natural radium concentration in surface waters Journal of Environmental Radioactivity 61(2) pp 27-37 2002

Lauria D C R M R Almeida O Sracek Behavior of radium thorium and uranium in groundwater near the Buena Lagoon in the Coastal Zone of the State of Rio de Janeiro Brazil Environmental Geology 47 pp 11-19 200

MACTEC 2005 Radiological Review of Fall 2004 RI Analytical Data Nuclear Metals Superfund Site Concord Massachusetts September 26 2004

MampE 2005 Memorandum from Andrew Schkuta to Ed Conroy Subject Uranium Analysis comparison of alpha spectrometry and ICPMS results December 15 2005

PSOP Project Summary and Operations Plan Nuclear Metals Superfund Site Concord Massachusetts April 15 2005

QAPP 2004 Quality Assurance Project Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Szabo Z D E Rice C L MacLeod and T H Barringer Relation of distribution of radium nitrate and pesticides to agricultural land use and depth Kirkwood-Cohansey aquifer system New Jersey Coastal Plain 1990-1991 US Geological Survey Water-Resources Investigations Report 96-4165A 119 p 1997

15

bull

bull bull

bull bull

bull bull bull

bull bull

bull bull

bull

500

1500

Figure 1

Con

cent

ratio

n (m

gkg

)

ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

1000

0

0 5 10 15 20 25 30 35 40

Samples

Figure 2

Alp

ha S

pect

osco

py (m

gkg

)

Fall 2004 Total Uranium No Super-Depleted

1000

100

10

1

y = 10604x R2 = 09049

1 10 100 1000

ICPMS (mgkg)

Fall 2004 Super-Depleted Linear Regression

Attachment 1Connolly Jayme

From Taylor Mel issagepam ail epa gov Sent Wednesday December 21 2005 932AM To brucetdemaximiscom jhuntdemaximiscom Subject Fw Uranium Analysis Issues

Rad-versus-ICPMS RADvsiCPMS_soilse MEvsPRP_ICPMSP memodoc (158 dPDF (29 KB) OF (25 KB)

brucejohn here is mampes take on the rns vs alpha spec analysis discrepancy

Forwarded by MelissaG TaylorR1USEPAUS on 12212005 0930 AM

Conroy Ed ltEdConroyregrn-ea ecomcomgt To

MelissaG TaylorR1USEPAUSregEPA 12152005 0416 cc PM

Subject FW Uranium Analysis Issues

Melissa

Heres Andys take on the alpha spec vs ICPMS debate along with a couple of graphs showing correlation between the two methods and MampE split data vs PRP data

Please call me on the cell phone tomorrow to discuss

Ed

Edward A Conroy PE

Project Manager

METCALF amp EDDYJAECOM

7812246880 ph

6172832942 cell

1

ed conroym-e aecom com Attachment 1

-----Original Message----shyFrom Schkuta Andrew Sent Thursday December 15 2005 358 PM To Conroy Ed Subject Uranium Analysis Issues

Ed

Please find attached a memorandum that summarizes my review of why ICP-MS and alpha spec data may not necessarily agree

ltltRad-versus-ICPMS memodocgtgt

In addition I have attached 2 preliminary charts that have the following draft information

ltltRADvsiCPMS soilsedPDFgtgt

The RADvsiCPMS soilsed chart is a draft plot of MampEs alpha spec versus ICP-MS results when data was available for both samples It shows the same general trend as the PRP data with fairly good agreement between the two analyses There appears to be a slight positive bias for alpha spec data at low concentrations but this is expected as discussed in my memo

ltltMEvsPRP ICPMSPDFgtgt

The MEvsPRP ICPMS is a draft plot of MampE vs the PRPs total uranium results as measured by ICP-MS -With the exception of a few outliers which will be further investigated there appears to be good agreement between MampEs and the PRPs data As a note the median percent difference between MampEs and the PRPs data was 1

Andrew Schkuta CHMM

Technical Specialist - Chemistry

METCALF amp EDDYJAECOM

701 Edgewater Drive

Wakefield MA 01880

(781) 224-6353

(781) 245-6293 fax

2

(See attached file Rad-versus-ICPMS Attachment 1memodoc) (See attached file RADvsiCPMS soilsedPDF) (See attached file MEvsPRP ICPMSPDF)

3

Attachment 1

Metcalf amp Eddy Inc

METCALFampEDDY

Memorandum

Date December 15 2005

To Ed Conroy

From Andrew Schkuta

Subject Uranium Analysis comparison of alpha spectrometry and ICP-MS results

File WA151 Nuclear Metals Distribution

Summary This memorandum discusses the differences between uranium analysis conducted by two different methods alpha spectroscopy (alpha spec) and inductively coupled plasma shymass spectrometry (ICP-MS) In summary the alpha spec data is expected to be biased high compared to the ICP-MS data but this is not indicative that either method is being performed incorrectly

In the September 25 2005 technical memorandum Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts (MACTEC 2005) data was presented comparing alpha spec and ICP-MS analysis for uranium analysis and it was proposed that future uranium analysis be performed by ICP-MS only For samples analyzed that were collected through Spring 2005 all samples were analyzed by ICP-MS with a subset (20) analyzed by alpha spec It was noted that there appeared to be a positive bias in the alpha spec data compared to the ICP-MS data and some concern has been voiced if analysis by ICP-MS only would meet project objectives This memorandum provides background on why the two analytical methods may yield differing results and discusses how they may impact project quality objectives

There are two primary reasons why alpha spec is expected to yield results that may be positively biased

1 Sample Preparation The sample preparation method for alpha spec is more rigorous The preparation method in alpha spec breaks down the silica mineral structure of the sample By design the ICP-MS sample preparation does not break down the mineral structure and is intended to remove all metals that could become environmentally available

2 Sample Quanitation The results for alpha spec are adjusted for analyte loss by correcting the result based on the amount of tracer recovered (yield) ICP-MS data is absolute ie data is not corrected for loss during sample preparation

2 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

These two reasons are discussed in further detail as follows

Sample Preparation

Radionuclide analysis traditionally uses a sample preparation method that includes hydrofluoric acid (HF) to break down the sample so that all of the uranium present in the sample is put into solution On the other hand standard EPA methods for metals analysis including ICP-MS do not measure the total metals present in a sample For example method SW846-3050 (EPA 1996) (as specified by the PRP in their approved QAPP for metals analysis by ICP-MS) states

This method is not a total digestion technique for most samples It is a very strong acid digestion that will dissolve almost all elements that could become environmentally available By design elements bound in silicate structures are not normally dissolved by this procedure as they are not usually mobile in the environment If absolute total digestion is required use Method 3052

As noted by the method a SW846 method is available to do a total digestion (equivalent to the alpha spec method) as method 3052 However method 3052 (EPA 1996) states (emphasis added)

Note This technique is not appropriate for regulatory applications that require the use of leachate preparations (ie Method 3050 Method 3051 Method 1311 Method 1312 Method 1310 Method 1320 Method 1330 Method 3031 Method 3040) This method is appropriate for those applications requiring a total decomposition for research purposes (ie~ geological studiesr mass balancesr analysis of Standard Reference Materials) or in response to a regulation that requires total sample decomposition

Even though the ICP-MS preparation method does not yield total results it still may conservatively measure the amount of metals that are environmentally available In the Research Project Summary Ambient Levels of Metals in New Jersey Soils prepared by the New Jersey Department of Environmental Protection (NJDEP 2003) where samples were prepared using SW846-3050 procedures it states

Metal concentrations reported in this summary were acid-extractable metals not necessarily total metal concentrations The extraction method (USEPA Method 3050) is a vigorous extraction method designed to remove all metals that could become environmentally available In practice the extraction method likely overestimates metals that could become available since the sample is refluxed with both concentrated nitric acid and hydrogen peroxide However by design the method will not extract chemicals incorporated in silica minerals as they are usually not mobile in the environment Thus concentrations reported in these studies may be lower than those indicated from analyses using methods designed to measure total metal concentrations such as x-ray fluorescence methods

Although references to uranium were not located it is expected that uranium will have some of the same type of mobility characteristics as other metals in the soil matrix The paper Ground Water Issue Behavior of Metals in Soil published by EPAs Superfund Technology Support Center for Ground Water (EPA 1992) describes how metals are distributed within soil The paper describes how metals are partitioned in various pools and how a rigorous extraction using HF is required to release metals present in the structure of primary or secondary minerals It is

3 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

expected that native metals (ie background concentrations) can be present in these structures The paper indicates that for four metals examined (cadmium copper zinc and nickel) the amount of metals that was bound in the mineral structure ranges from approximately 15-60 It would be expected that naturally occurring uranium would have similar characteristics

The National Institutes of Standards amp Technology (NIST) recognizes the differences between total digestion and method SW846-3050 and has issued Addendum to Certificates for the Certificate of Analysis (NIST 2003) for some Standard Reference Materials (SRMs) The Addendum to Certificates for SRM 2711 a Montana agricultural soil indicates that trace metals were measured at approximately 40 to 90 of their certified value when analyzed using SW846-3050 as the preparation method Even though uranium is not listed as a target analyte it is reasonable to expect that is would behave in a similar fashion

Sample Quanitation

The preparation method for alpha spec analysis includes addition of a tracer (U-232) to sample and is carried through the entire analytical scheme The amount of tracer is measured during analysis and the results for target compounds (the other uranium isotopes) are corrected for the amount of tracer recovery (yield) For example if amount of tracer measured is 50 of what was initially added to the sample the sample results are all corrected by dividing by 050 In terms of method quality assurancequality control (QAQC) the PRP QAPP indicates that tracer recovery in the range of 15-125 is considered acceptable and the analytical system is considered in control In other words up to 85 of the compound can be lost during sample preparation and analysis and the data is considered meeting all QAQC objectives This method of analysis is the equivalent to the technique of isotope dilution which is used for some non-radiological environmental analyses (eg EPA Method 1625 Semivolatile Organic Compounds by Isotope Dilution GCMS) Analyses using this technique although considered the most accurate are not widely employed due to the complexity and expense of using the isotope dilution technique

ICP-MS analysis measures the concentration of analyte with no correction for potential losses However QAQC procedures are in place to make sure the losses are not considered significant In addition to laboratory control samples matrix spike analysis measures the recovery for the target analyte that was over-spiked into the sample The PRP QAPP states that matrix spike recovery should be within the range of 75-125 Therefore up to 25 of the analyte could be lost and the method has still met the QAQC objectives If recovery is less than this then validation procedures acknowledge that the data may be biased low and sample results are flagged as estimated

Therefore because the alpha spec data is corrected for analyte loss during the analytical process while ICP-MS data is not some amount of positive bias is expected to be present in alpha spec data when compared to ICP-MS data

Considerations

In evaluating which analysis is appropriate for samples from the project neither method is inherently better than the other The following factors should be considered

1 What is being measured ICP-MS measures mass directly and data can be mathematically converted to activity Alpha spec measures activity directly and can be mathematically converted to mass

4 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

2 What are the objectives of the data in terms of measuring total versus environmentally available The ICP-MS method specified in the PRP QAPP measures uranium concentrations in the same manner that EPA has measured other metals concentrations Should uranium mass be measured in a typical way that will be used for measuring other metals of concern at the site (eg copper beryllium arsenic etc)

3 If ICP-MS is more cost effective that alpha spec should the option of using preparation method SW846-3052 be considered This should yield total uranium data similar to alpha spec analysis while still using the efficiencies of ICP-MS analysis

4 What units will the cleanup level for the site be in If its in mass units then ICPshyMS analysis may be more appropriate If its in activity units then alpha spec analysis may be warranted

References

Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts Technical Memorandum from Heath Downey to Bruce Thompson de maxim is inc September 26 2005 (MACTEC 2005)

Test Methods for Evaluating Solid Waste PhysicalChemical Methods SW-846 Third Edition with updates EPA 1996

Ambient Levels of Metals in New Jersey Soils Environmental Assessment and Risk Analysis Element Research Project Summary Division of Science Research and Technology State of New Jersey Department of Environmental Protection May2003

Ground Water Issue Behavior of Metals in Soil EPA Superfund Technology Support Center for Ground Water EPA540S-92018 October 1993

Certificate of Analysis Standard Reference Material 2711 Montana Soil National Institute of Standards amp Technology July 18 2003

bull bull

bull bull bullbull bull

10000

MampE SoilSediment Results Total Uranium mgKg

bull

bull - ~-~-----~~---~~-~--~~1000

en aE

I I bull Total a u

I -

bull bull 100 1shy -middot- -middot bullbull bull

Ybull bull

10 100 1000 10000

Alpha Spec

Attachm

ent 1

10 -

100000

TOTAL URANIUM BY ICP-MS mgKg

bull

bull 10000

I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 14: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

de maximis inc

Figure 1 ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

Figure 2 Comparison of Alpha Spectroscopy and ICP-MS Uranium Analysis

Attachment 1 Metcalf amp Eddy Memorandum ndash Uranium Analysis comparison of alpha spectrometry and ICP-MS results

Attachment 2 QAPP SOPs for Alpha Spectroscopy and ICPMS Sample Preparation

Attachment 3 Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor Department of the Army January 19 2000

References

Beneš P Migration of radium in the terrestrial hydrosphere The behavior of radium in waterways and aquifers Technical Reports Series No 310 International Atomic Energy Agency IAEA-TECDOC-301 Vienna Austrai pp 119-173 1984

Beneš P M Obdržaacutelek M Čejchanovaacute The physicochemical forms of tracers of radium in aqueous solutions containing chlorides sulfates and carbonates Radiochem Radioanal Letters 50(4) pp 227-242 1982

CREW 2006 ldquoComments on Radiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Technical Note No 2006-01 January 6 2006

Demir I Temporal variations and sources of Pb Co Cu Ni Fe and Mo in shallow ground water of the McDonalds Branch basin Lebanon State Forest Burlington County New Jersey New Jersey Geological Survey Report 20 19 p 1989

EPA 2000 Guidance for Data Quality Assessment Practical Methods for Data Analysis EPA QAG-9 July 2000

EPA 2002 Guidance for Comparing Background and Chemical Concentrations in Soil for CERCLA Site EPA 540-R-01-003 September 2002

FSP 2004 Field Sampling Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Gilbert 1989 Statistical Methods for Environmental Pollution Monitoring Richard O Gilbert 1987

Kirby H W M L Salutsky The Radiochemistry of Radium National Academy of Sciences ndash National Research Council NAS-NS 3057 205 p 1964

14

de maximis inc

Kramer T F and D F Reid The occurrence and behavior of radium in saline formation water of the US Gulf Coast Region Isotope Geoscience 2 pp 153shy174 1984

Landreacute V Comments on ldquoRadiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Nuclear Metals Superfund Site CREW Technical Note 2006-01 January 6 2006

Lauria D C and J M O Godoy Abnormal high natural radium concentration in surface waters Journal of Environmental Radioactivity 61(2) pp 27-37 2002

Lauria D C R M R Almeida O Sracek Behavior of radium thorium and uranium in groundwater near the Buena Lagoon in the Coastal Zone of the State of Rio de Janeiro Brazil Environmental Geology 47 pp 11-19 200

MACTEC 2005 Radiological Review of Fall 2004 RI Analytical Data Nuclear Metals Superfund Site Concord Massachusetts September 26 2004

MampE 2005 Memorandum from Andrew Schkuta to Ed Conroy Subject Uranium Analysis comparison of alpha spectrometry and ICPMS results December 15 2005

PSOP Project Summary and Operations Plan Nuclear Metals Superfund Site Concord Massachusetts April 15 2005

QAPP 2004 Quality Assurance Project Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Szabo Z D E Rice C L MacLeod and T H Barringer Relation of distribution of radium nitrate and pesticides to agricultural land use and depth Kirkwood-Cohansey aquifer system New Jersey Coastal Plain 1990-1991 US Geological Survey Water-Resources Investigations Report 96-4165A 119 p 1997

15

bull

bull bull

bull bull

bull bull bull

bull bull

bull bull

bull

500

1500

Figure 1

Con

cent

ratio

n (m

gkg

)

ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

1000

0

0 5 10 15 20 25 30 35 40

Samples

Figure 2

Alp

ha S

pect

osco

py (m

gkg

)

Fall 2004 Total Uranium No Super-Depleted

1000

100

10

1

y = 10604x R2 = 09049

1 10 100 1000

ICPMS (mgkg)

Fall 2004 Super-Depleted Linear Regression

Attachment 1Connolly Jayme

From Taylor Mel issagepam ail epa gov Sent Wednesday December 21 2005 932AM To brucetdemaximiscom jhuntdemaximiscom Subject Fw Uranium Analysis Issues

Rad-versus-ICPMS RADvsiCPMS_soilse MEvsPRP_ICPMSP memodoc (158 dPDF (29 KB) OF (25 KB)

brucejohn here is mampes take on the rns vs alpha spec analysis discrepancy

Forwarded by MelissaG TaylorR1USEPAUS on 12212005 0930 AM

Conroy Ed ltEdConroyregrn-ea ecomcomgt To

MelissaG TaylorR1USEPAUSregEPA 12152005 0416 cc PM

Subject FW Uranium Analysis Issues

Melissa

Heres Andys take on the alpha spec vs ICPMS debate along with a couple of graphs showing correlation between the two methods and MampE split data vs PRP data

Please call me on the cell phone tomorrow to discuss

Ed

Edward A Conroy PE

Project Manager

METCALF amp EDDYJAECOM

7812246880 ph

6172832942 cell

1

ed conroym-e aecom com Attachment 1

-----Original Message----shyFrom Schkuta Andrew Sent Thursday December 15 2005 358 PM To Conroy Ed Subject Uranium Analysis Issues

Ed

Please find attached a memorandum that summarizes my review of why ICP-MS and alpha spec data may not necessarily agree

ltltRad-versus-ICPMS memodocgtgt

In addition I have attached 2 preliminary charts that have the following draft information

ltltRADvsiCPMS soilsedPDFgtgt

The RADvsiCPMS soilsed chart is a draft plot of MampEs alpha spec versus ICP-MS results when data was available for both samples It shows the same general trend as the PRP data with fairly good agreement between the two analyses There appears to be a slight positive bias for alpha spec data at low concentrations but this is expected as discussed in my memo

ltltMEvsPRP ICPMSPDFgtgt

The MEvsPRP ICPMS is a draft plot of MampE vs the PRPs total uranium results as measured by ICP-MS -With the exception of a few outliers which will be further investigated there appears to be good agreement between MampEs and the PRPs data As a note the median percent difference between MampEs and the PRPs data was 1

Andrew Schkuta CHMM

Technical Specialist - Chemistry

METCALF amp EDDYJAECOM

701 Edgewater Drive

Wakefield MA 01880

(781) 224-6353

(781) 245-6293 fax

2

(See attached file Rad-versus-ICPMS Attachment 1memodoc) (See attached file RADvsiCPMS soilsedPDF) (See attached file MEvsPRP ICPMSPDF)

3

Attachment 1

Metcalf amp Eddy Inc

METCALFampEDDY

Memorandum

Date December 15 2005

To Ed Conroy

From Andrew Schkuta

Subject Uranium Analysis comparison of alpha spectrometry and ICP-MS results

File WA151 Nuclear Metals Distribution

Summary This memorandum discusses the differences between uranium analysis conducted by two different methods alpha spectroscopy (alpha spec) and inductively coupled plasma shymass spectrometry (ICP-MS) In summary the alpha spec data is expected to be biased high compared to the ICP-MS data but this is not indicative that either method is being performed incorrectly

In the September 25 2005 technical memorandum Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts (MACTEC 2005) data was presented comparing alpha spec and ICP-MS analysis for uranium analysis and it was proposed that future uranium analysis be performed by ICP-MS only For samples analyzed that were collected through Spring 2005 all samples were analyzed by ICP-MS with a subset (20) analyzed by alpha spec It was noted that there appeared to be a positive bias in the alpha spec data compared to the ICP-MS data and some concern has been voiced if analysis by ICP-MS only would meet project objectives This memorandum provides background on why the two analytical methods may yield differing results and discusses how they may impact project quality objectives

There are two primary reasons why alpha spec is expected to yield results that may be positively biased

1 Sample Preparation The sample preparation method for alpha spec is more rigorous The preparation method in alpha spec breaks down the silica mineral structure of the sample By design the ICP-MS sample preparation does not break down the mineral structure and is intended to remove all metals that could become environmentally available

2 Sample Quanitation The results for alpha spec are adjusted for analyte loss by correcting the result based on the amount of tracer recovered (yield) ICP-MS data is absolute ie data is not corrected for loss during sample preparation

2 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

These two reasons are discussed in further detail as follows

Sample Preparation

Radionuclide analysis traditionally uses a sample preparation method that includes hydrofluoric acid (HF) to break down the sample so that all of the uranium present in the sample is put into solution On the other hand standard EPA methods for metals analysis including ICP-MS do not measure the total metals present in a sample For example method SW846-3050 (EPA 1996) (as specified by the PRP in their approved QAPP for metals analysis by ICP-MS) states

This method is not a total digestion technique for most samples It is a very strong acid digestion that will dissolve almost all elements that could become environmentally available By design elements bound in silicate structures are not normally dissolved by this procedure as they are not usually mobile in the environment If absolute total digestion is required use Method 3052

As noted by the method a SW846 method is available to do a total digestion (equivalent to the alpha spec method) as method 3052 However method 3052 (EPA 1996) states (emphasis added)

Note This technique is not appropriate for regulatory applications that require the use of leachate preparations (ie Method 3050 Method 3051 Method 1311 Method 1312 Method 1310 Method 1320 Method 1330 Method 3031 Method 3040) This method is appropriate for those applications requiring a total decomposition for research purposes (ie~ geological studiesr mass balancesr analysis of Standard Reference Materials) or in response to a regulation that requires total sample decomposition

Even though the ICP-MS preparation method does not yield total results it still may conservatively measure the amount of metals that are environmentally available In the Research Project Summary Ambient Levels of Metals in New Jersey Soils prepared by the New Jersey Department of Environmental Protection (NJDEP 2003) where samples were prepared using SW846-3050 procedures it states

Metal concentrations reported in this summary were acid-extractable metals not necessarily total metal concentrations The extraction method (USEPA Method 3050) is a vigorous extraction method designed to remove all metals that could become environmentally available In practice the extraction method likely overestimates metals that could become available since the sample is refluxed with both concentrated nitric acid and hydrogen peroxide However by design the method will not extract chemicals incorporated in silica minerals as they are usually not mobile in the environment Thus concentrations reported in these studies may be lower than those indicated from analyses using methods designed to measure total metal concentrations such as x-ray fluorescence methods

Although references to uranium were not located it is expected that uranium will have some of the same type of mobility characteristics as other metals in the soil matrix The paper Ground Water Issue Behavior of Metals in Soil published by EPAs Superfund Technology Support Center for Ground Water (EPA 1992) describes how metals are distributed within soil The paper describes how metals are partitioned in various pools and how a rigorous extraction using HF is required to release metals present in the structure of primary or secondary minerals It is

3 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

expected that native metals (ie background concentrations) can be present in these structures The paper indicates that for four metals examined (cadmium copper zinc and nickel) the amount of metals that was bound in the mineral structure ranges from approximately 15-60 It would be expected that naturally occurring uranium would have similar characteristics

The National Institutes of Standards amp Technology (NIST) recognizes the differences between total digestion and method SW846-3050 and has issued Addendum to Certificates for the Certificate of Analysis (NIST 2003) for some Standard Reference Materials (SRMs) The Addendum to Certificates for SRM 2711 a Montana agricultural soil indicates that trace metals were measured at approximately 40 to 90 of their certified value when analyzed using SW846-3050 as the preparation method Even though uranium is not listed as a target analyte it is reasonable to expect that is would behave in a similar fashion

Sample Quanitation

The preparation method for alpha spec analysis includes addition of a tracer (U-232) to sample and is carried through the entire analytical scheme The amount of tracer is measured during analysis and the results for target compounds (the other uranium isotopes) are corrected for the amount of tracer recovery (yield) For example if amount of tracer measured is 50 of what was initially added to the sample the sample results are all corrected by dividing by 050 In terms of method quality assurancequality control (QAQC) the PRP QAPP indicates that tracer recovery in the range of 15-125 is considered acceptable and the analytical system is considered in control In other words up to 85 of the compound can be lost during sample preparation and analysis and the data is considered meeting all QAQC objectives This method of analysis is the equivalent to the technique of isotope dilution which is used for some non-radiological environmental analyses (eg EPA Method 1625 Semivolatile Organic Compounds by Isotope Dilution GCMS) Analyses using this technique although considered the most accurate are not widely employed due to the complexity and expense of using the isotope dilution technique

ICP-MS analysis measures the concentration of analyte with no correction for potential losses However QAQC procedures are in place to make sure the losses are not considered significant In addition to laboratory control samples matrix spike analysis measures the recovery for the target analyte that was over-spiked into the sample The PRP QAPP states that matrix spike recovery should be within the range of 75-125 Therefore up to 25 of the analyte could be lost and the method has still met the QAQC objectives If recovery is less than this then validation procedures acknowledge that the data may be biased low and sample results are flagged as estimated

Therefore because the alpha spec data is corrected for analyte loss during the analytical process while ICP-MS data is not some amount of positive bias is expected to be present in alpha spec data when compared to ICP-MS data

Considerations

In evaluating which analysis is appropriate for samples from the project neither method is inherently better than the other The following factors should be considered

1 What is being measured ICP-MS measures mass directly and data can be mathematically converted to activity Alpha spec measures activity directly and can be mathematically converted to mass

4 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

2 What are the objectives of the data in terms of measuring total versus environmentally available The ICP-MS method specified in the PRP QAPP measures uranium concentrations in the same manner that EPA has measured other metals concentrations Should uranium mass be measured in a typical way that will be used for measuring other metals of concern at the site (eg copper beryllium arsenic etc)

3 If ICP-MS is more cost effective that alpha spec should the option of using preparation method SW846-3052 be considered This should yield total uranium data similar to alpha spec analysis while still using the efficiencies of ICP-MS analysis

4 What units will the cleanup level for the site be in If its in mass units then ICPshyMS analysis may be more appropriate If its in activity units then alpha spec analysis may be warranted

References

Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts Technical Memorandum from Heath Downey to Bruce Thompson de maxim is inc September 26 2005 (MACTEC 2005)

Test Methods for Evaluating Solid Waste PhysicalChemical Methods SW-846 Third Edition with updates EPA 1996

Ambient Levels of Metals in New Jersey Soils Environmental Assessment and Risk Analysis Element Research Project Summary Division of Science Research and Technology State of New Jersey Department of Environmental Protection May2003

Ground Water Issue Behavior of Metals in Soil EPA Superfund Technology Support Center for Ground Water EPA540S-92018 October 1993

Certificate of Analysis Standard Reference Material 2711 Montana Soil National Institute of Standards amp Technology July 18 2003

bull bull

bull bull bullbull bull

10000

MampE SoilSediment Results Total Uranium mgKg

bull

bull - ~-~-----~~---~~-~--~~1000

en aE

I I bull Total a u

I -

bull bull 100 1shy -middot- -middot bullbull bull

Ybull bull

10 100 1000 10000

Alpha Spec

Attachm

ent 1

10 -

100000

TOTAL URANIUM BY ICP-MS mgKg

bull

bull 10000

I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 15: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

de maximis inc

Kramer T F and D F Reid The occurrence and behavior of radium in saline formation water of the US Gulf Coast Region Isotope Geoscience 2 pp 153shy174 1984

Landreacute V Comments on ldquoRadiological Review of Fall 2004 RI Analytical Data September 26 2005rdquo Nuclear Metals Superfund Site CREW Technical Note 2006-01 January 6 2006

Lauria D C and J M O Godoy Abnormal high natural radium concentration in surface waters Journal of Environmental Radioactivity 61(2) pp 27-37 2002

Lauria D C R M R Almeida O Sracek Behavior of radium thorium and uranium in groundwater near the Buena Lagoon in the Coastal Zone of the State of Rio de Janeiro Brazil Environmental Geology 47 pp 11-19 200

MACTEC 2005 Radiological Review of Fall 2004 RI Analytical Data Nuclear Metals Superfund Site Concord Massachusetts September 26 2004

MampE 2005 Memorandum from Andrew Schkuta to Ed Conroy Subject Uranium Analysis comparison of alpha spectrometry and ICPMS results December 15 2005

PSOP Project Summary and Operations Plan Nuclear Metals Superfund Site Concord Massachusetts April 15 2005

QAPP 2004 Quality Assurance Project Plan Nuclear Metals Superfund Site Concord Massachusetts September 29 2004

Szabo Z D E Rice C L MacLeod and T H Barringer Relation of distribution of radium nitrate and pesticides to agricultural land use and depth Kirkwood-Cohansey aquifer system New Jersey Coastal Plain 1990-1991 US Geological Survey Water-Resources Investigations Report 96-4165A 119 p 1997

15

bull

bull bull

bull bull

bull bull bull

bull bull

bull bull

bull

500

1500

Figure 1

Con

cent

ratio

n (m

gkg

)

ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

1000

0

0 5 10 15 20 25 30 35 40

Samples

Figure 2

Alp

ha S

pect

osco

py (m

gkg

)

Fall 2004 Total Uranium No Super-Depleted

1000

100

10

1

y = 10604x R2 = 09049

1 10 100 1000

ICPMS (mgkg)

Fall 2004 Super-Depleted Linear Regression

Attachment 1Connolly Jayme

From Taylor Mel issagepam ail epa gov Sent Wednesday December 21 2005 932AM To brucetdemaximiscom jhuntdemaximiscom Subject Fw Uranium Analysis Issues

Rad-versus-ICPMS RADvsiCPMS_soilse MEvsPRP_ICPMSP memodoc (158 dPDF (29 KB) OF (25 KB)

brucejohn here is mampes take on the rns vs alpha spec analysis discrepancy

Forwarded by MelissaG TaylorR1USEPAUS on 12212005 0930 AM

Conroy Ed ltEdConroyregrn-ea ecomcomgt To

MelissaG TaylorR1USEPAUSregEPA 12152005 0416 cc PM

Subject FW Uranium Analysis Issues

Melissa

Heres Andys take on the alpha spec vs ICPMS debate along with a couple of graphs showing correlation between the two methods and MampE split data vs PRP data

Please call me on the cell phone tomorrow to discuss

Ed

Edward A Conroy PE

Project Manager

METCALF amp EDDYJAECOM

7812246880 ph

6172832942 cell

1

ed conroym-e aecom com Attachment 1

-----Original Message----shyFrom Schkuta Andrew Sent Thursday December 15 2005 358 PM To Conroy Ed Subject Uranium Analysis Issues

Ed

Please find attached a memorandum that summarizes my review of why ICP-MS and alpha spec data may not necessarily agree

ltltRad-versus-ICPMS memodocgtgt

In addition I have attached 2 preliminary charts that have the following draft information

ltltRADvsiCPMS soilsedPDFgtgt

The RADvsiCPMS soilsed chart is a draft plot of MampEs alpha spec versus ICP-MS results when data was available for both samples It shows the same general trend as the PRP data with fairly good agreement between the two analyses There appears to be a slight positive bias for alpha spec data at low concentrations but this is expected as discussed in my memo

ltltMEvsPRP ICPMSPDFgtgt

The MEvsPRP ICPMS is a draft plot of MampE vs the PRPs total uranium results as measured by ICP-MS -With the exception of a few outliers which will be further investigated there appears to be good agreement between MampEs and the PRPs data As a note the median percent difference between MampEs and the PRPs data was 1

Andrew Schkuta CHMM

Technical Specialist - Chemistry

METCALF amp EDDYJAECOM

701 Edgewater Drive

Wakefield MA 01880

(781) 224-6353

(781) 245-6293 fax

2

(See attached file Rad-versus-ICPMS Attachment 1memodoc) (See attached file RADvsiCPMS soilsedPDF) (See attached file MEvsPRP ICPMSPDF)

3

Attachment 1

Metcalf amp Eddy Inc

METCALFampEDDY

Memorandum

Date December 15 2005

To Ed Conroy

From Andrew Schkuta

Subject Uranium Analysis comparison of alpha spectrometry and ICP-MS results

File WA151 Nuclear Metals Distribution

Summary This memorandum discusses the differences between uranium analysis conducted by two different methods alpha spectroscopy (alpha spec) and inductively coupled plasma shymass spectrometry (ICP-MS) In summary the alpha spec data is expected to be biased high compared to the ICP-MS data but this is not indicative that either method is being performed incorrectly

In the September 25 2005 technical memorandum Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts (MACTEC 2005) data was presented comparing alpha spec and ICP-MS analysis for uranium analysis and it was proposed that future uranium analysis be performed by ICP-MS only For samples analyzed that were collected through Spring 2005 all samples were analyzed by ICP-MS with a subset (20) analyzed by alpha spec It was noted that there appeared to be a positive bias in the alpha spec data compared to the ICP-MS data and some concern has been voiced if analysis by ICP-MS only would meet project objectives This memorandum provides background on why the two analytical methods may yield differing results and discusses how they may impact project quality objectives

There are two primary reasons why alpha spec is expected to yield results that may be positively biased

1 Sample Preparation The sample preparation method for alpha spec is more rigorous The preparation method in alpha spec breaks down the silica mineral structure of the sample By design the ICP-MS sample preparation does not break down the mineral structure and is intended to remove all metals that could become environmentally available

2 Sample Quanitation The results for alpha spec are adjusted for analyte loss by correcting the result based on the amount of tracer recovered (yield) ICP-MS data is absolute ie data is not corrected for loss during sample preparation

2 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

These two reasons are discussed in further detail as follows

Sample Preparation

Radionuclide analysis traditionally uses a sample preparation method that includes hydrofluoric acid (HF) to break down the sample so that all of the uranium present in the sample is put into solution On the other hand standard EPA methods for metals analysis including ICP-MS do not measure the total metals present in a sample For example method SW846-3050 (EPA 1996) (as specified by the PRP in their approved QAPP for metals analysis by ICP-MS) states

This method is not a total digestion technique for most samples It is a very strong acid digestion that will dissolve almost all elements that could become environmentally available By design elements bound in silicate structures are not normally dissolved by this procedure as they are not usually mobile in the environment If absolute total digestion is required use Method 3052

As noted by the method a SW846 method is available to do a total digestion (equivalent to the alpha spec method) as method 3052 However method 3052 (EPA 1996) states (emphasis added)

Note This technique is not appropriate for regulatory applications that require the use of leachate preparations (ie Method 3050 Method 3051 Method 1311 Method 1312 Method 1310 Method 1320 Method 1330 Method 3031 Method 3040) This method is appropriate for those applications requiring a total decomposition for research purposes (ie~ geological studiesr mass balancesr analysis of Standard Reference Materials) or in response to a regulation that requires total sample decomposition

Even though the ICP-MS preparation method does not yield total results it still may conservatively measure the amount of metals that are environmentally available In the Research Project Summary Ambient Levels of Metals in New Jersey Soils prepared by the New Jersey Department of Environmental Protection (NJDEP 2003) where samples were prepared using SW846-3050 procedures it states

Metal concentrations reported in this summary were acid-extractable metals not necessarily total metal concentrations The extraction method (USEPA Method 3050) is a vigorous extraction method designed to remove all metals that could become environmentally available In practice the extraction method likely overestimates metals that could become available since the sample is refluxed with both concentrated nitric acid and hydrogen peroxide However by design the method will not extract chemicals incorporated in silica minerals as they are usually not mobile in the environment Thus concentrations reported in these studies may be lower than those indicated from analyses using methods designed to measure total metal concentrations such as x-ray fluorescence methods

Although references to uranium were not located it is expected that uranium will have some of the same type of mobility characteristics as other metals in the soil matrix The paper Ground Water Issue Behavior of Metals in Soil published by EPAs Superfund Technology Support Center for Ground Water (EPA 1992) describes how metals are distributed within soil The paper describes how metals are partitioned in various pools and how a rigorous extraction using HF is required to release metals present in the structure of primary or secondary minerals It is

3 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

expected that native metals (ie background concentrations) can be present in these structures The paper indicates that for four metals examined (cadmium copper zinc and nickel) the amount of metals that was bound in the mineral structure ranges from approximately 15-60 It would be expected that naturally occurring uranium would have similar characteristics

The National Institutes of Standards amp Technology (NIST) recognizes the differences between total digestion and method SW846-3050 and has issued Addendum to Certificates for the Certificate of Analysis (NIST 2003) for some Standard Reference Materials (SRMs) The Addendum to Certificates for SRM 2711 a Montana agricultural soil indicates that trace metals were measured at approximately 40 to 90 of their certified value when analyzed using SW846-3050 as the preparation method Even though uranium is not listed as a target analyte it is reasonable to expect that is would behave in a similar fashion

Sample Quanitation

The preparation method for alpha spec analysis includes addition of a tracer (U-232) to sample and is carried through the entire analytical scheme The amount of tracer is measured during analysis and the results for target compounds (the other uranium isotopes) are corrected for the amount of tracer recovery (yield) For example if amount of tracer measured is 50 of what was initially added to the sample the sample results are all corrected by dividing by 050 In terms of method quality assurancequality control (QAQC) the PRP QAPP indicates that tracer recovery in the range of 15-125 is considered acceptable and the analytical system is considered in control In other words up to 85 of the compound can be lost during sample preparation and analysis and the data is considered meeting all QAQC objectives This method of analysis is the equivalent to the technique of isotope dilution which is used for some non-radiological environmental analyses (eg EPA Method 1625 Semivolatile Organic Compounds by Isotope Dilution GCMS) Analyses using this technique although considered the most accurate are not widely employed due to the complexity and expense of using the isotope dilution technique

ICP-MS analysis measures the concentration of analyte with no correction for potential losses However QAQC procedures are in place to make sure the losses are not considered significant In addition to laboratory control samples matrix spike analysis measures the recovery for the target analyte that was over-spiked into the sample The PRP QAPP states that matrix spike recovery should be within the range of 75-125 Therefore up to 25 of the analyte could be lost and the method has still met the QAQC objectives If recovery is less than this then validation procedures acknowledge that the data may be biased low and sample results are flagged as estimated

Therefore because the alpha spec data is corrected for analyte loss during the analytical process while ICP-MS data is not some amount of positive bias is expected to be present in alpha spec data when compared to ICP-MS data

Considerations

In evaluating which analysis is appropriate for samples from the project neither method is inherently better than the other The following factors should be considered

1 What is being measured ICP-MS measures mass directly and data can be mathematically converted to activity Alpha spec measures activity directly and can be mathematically converted to mass

4 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

2 What are the objectives of the data in terms of measuring total versus environmentally available The ICP-MS method specified in the PRP QAPP measures uranium concentrations in the same manner that EPA has measured other metals concentrations Should uranium mass be measured in a typical way that will be used for measuring other metals of concern at the site (eg copper beryllium arsenic etc)

3 If ICP-MS is more cost effective that alpha spec should the option of using preparation method SW846-3052 be considered This should yield total uranium data similar to alpha spec analysis while still using the efficiencies of ICP-MS analysis

4 What units will the cleanup level for the site be in If its in mass units then ICPshyMS analysis may be more appropriate If its in activity units then alpha spec analysis may be warranted

References

Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts Technical Memorandum from Heath Downey to Bruce Thompson de maxim is inc September 26 2005 (MACTEC 2005)

Test Methods for Evaluating Solid Waste PhysicalChemical Methods SW-846 Third Edition with updates EPA 1996

Ambient Levels of Metals in New Jersey Soils Environmental Assessment and Risk Analysis Element Research Project Summary Division of Science Research and Technology State of New Jersey Department of Environmental Protection May2003

Ground Water Issue Behavior of Metals in Soil EPA Superfund Technology Support Center for Ground Water EPA540S-92018 October 1993

Certificate of Analysis Standard Reference Material 2711 Montana Soil National Institute of Standards amp Technology July 18 2003

bull bull

bull bull bullbull bull

10000

MampE SoilSediment Results Total Uranium mgKg

bull

bull - ~-~-----~~---~~-~--~~1000

en aE

I I bull Total a u

I -

bull bull 100 1shy -middot- -middot bullbull bull

Ybull bull

10 100 1000 10000

Alpha Spec

Attachm

ent 1

10 -

100000

TOTAL URANIUM BY ICP-MS mgKg

bull

bull 10000

I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 16: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

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500

1500

Figure 1

Con

cent

ratio

n (m

gkg

)

ICPMS Uranium Concentrations SDGs NMI-024 and NMI-027

1000

0

0 5 10 15 20 25 30 35 40

Samples

Figure 2

Alp

ha S

pect

osco

py (m

gkg

)

Fall 2004 Total Uranium No Super-Depleted

1000

100

10

1

y = 10604x R2 = 09049

1 10 100 1000

ICPMS (mgkg)

Fall 2004 Super-Depleted Linear Regression

Attachment 1Connolly Jayme

From Taylor Mel issagepam ail epa gov Sent Wednesday December 21 2005 932AM To brucetdemaximiscom jhuntdemaximiscom Subject Fw Uranium Analysis Issues

Rad-versus-ICPMS RADvsiCPMS_soilse MEvsPRP_ICPMSP memodoc (158 dPDF (29 KB) OF (25 KB)

brucejohn here is mampes take on the rns vs alpha spec analysis discrepancy

Forwarded by MelissaG TaylorR1USEPAUS on 12212005 0930 AM

Conroy Ed ltEdConroyregrn-ea ecomcomgt To

MelissaG TaylorR1USEPAUSregEPA 12152005 0416 cc PM

Subject FW Uranium Analysis Issues

Melissa

Heres Andys take on the alpha spec vs ICPMS debate along with a couple of graphs showing correlation between the two methods and MampE split data vs PRP data

Please call me on the cell phone tomorrow to discuss

Ed

Edward A Conroy PE

Project Manager

METCALF amp EDDYJAECOM

7812246880 ph

6172832942 cell

1

ed conroym-e aecom com Attachment 1

-----Original Message----shyFrom Schkuta Andrew Sent Thursday December 15 2005 358 PM To Conroy Ed Subject Uranium Analysis Issues

Ed

Please find attached a memorandum that summarizes my review of why ICP-MS and alpha spec data may not necessarily agree

ltltRad-versus-ICPMS memodocgtgt

In addition I have attached 2 preliminary charts that have the following draft information

ltltRADvsiCPMS soilsedPDFgtgt

The RADvsiCPMS soilsed chart is a draft plot of MampEs alpha spec versus ICP-MS results when data was available for both samples It shows the same general trend as the PRP data with fairly good agreement between the two analyses There appears to be a slight positive bias for alpha spec data at low concentrations but this is expected as discussed in my memo

ltltMEvsPRP ICPMSPDFgtgt

The MEvsPRP ICPMS is a draft plot of MampE vs the PRPs total uranium results as measured by ICP-MS -With the exception of a few outliers which will be further investigated there appears to be good agreement between MampEs and the PRPs data As a note the median percent difference between MampEs and the PRPs data was 1

Andrew Schkuta CHMM

Technical Specialist - Chemistry

METCALF amp EDDYJAECOM

701 Edgewater Drive

Wakefield MA 01880

(781) 224-6353

(781) 245-6293 fax

2

(See attached file Rad-versus-ICPMS Attachment 1memodoc) (See attached file RADvsiCPMS soilsedPDF) (See attached file MEvsPRP ICPMSPDF)

3

Attachment 1

Metcalf amp Eddy Inc

METCALFampEDDY

Memorandum

Date December 15 2005

To Ed Conroy

From Andrew Schkuta

Subject Uranium Analysis comparison of alpha spectrometry and ICP-MS results

File WA151 Nuclear Metals Distribution

Summary This memorandum discusses the differences between uranium analysis conducted by two different methods alpha spectroscopy (alpha spec) and inductively coupled plasma shymass spectrometry (ICP-MS) In summary the alpha spec data is expected to be biased high compared to the ICP-MS data but this is not indicative that either method is being performed incorrectly

In the September 25 2005 technical memorandum Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts (MACTEC 2005) data was presented comparing alpha spec and ICP-MS analysis for uranium analysis and it was proposed that future uranium analysis be performed by ICP-MS only For samples analyzed that were collected through Spring 2005 all samples were analyzed by ICP-MS with a subset (20) analyzed by alpha spec It was noted that there appeared to be a positive bias in the alpha spec data compared to the ICP-MS data and some concern has been voiced if analysis by ICP-MS only would meet project objectives This memorandum provides background on why the two analytical methods may yield differing results and discusses how they may impact project quality objectives

There are two primary reasons why alpha spec is expected to yield results that may be positively biased

1 Sample Preparation The sample preparation method for alpha spec is more rigorous The preparation method in alpha spec breaks down the silica mineral structure of the sample By design the ICP-MS sample preparation does not break down the mineral structure and is intended to remove all metals that could become environmentally available

2 Sample Quanitation The results for alpha spec are adjusted for analyte loss by correcting the result based on the amount of tracer recovered (yield) ICP-MS data is absolute ie data is not corrected for loss during sample preparation

2 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

These two reasons are discussed in further detail as follows

Sample Preparation

Radionuclide analysis traditionally uses a sample preparation method that includes hydrofluoric acid (HF) to break down the sample so that all of the uranium present in the sample is put into solution On the other hand standard EPA methods for metals analysis including ICP-MS do not measure the total metals present in a sample For example method SW846-3050 (EPA 1996) (as specified by the PRP in their approved QAPP for metals analysis by ICP-MS) states

This method is not a total digestion technique for most samples It is a very strong acid digestion that will dissolve almost all elements that could become environmentally available By design elements bound in silicate structures are not normally dissolved by this procedure as they are not usually mobile in the environment If absolute total digestion is required use Method 3052

As noted by the method a SW846 method is available to do a total digestion (equivalent to the alpha spec method) as method 3052 However method 3052 (EPA 1996) states (emphasis added)

Note This technique is not appropriate for regulatory applications that require the use of leachate preparations (ie Method 3050 Method 3051 Method 1311 Method 1312 Method 1310 Method 1320 Method 1330 Method 3031 Method 3040) This method is appropriate for those applications requiring a total decomposition for research purposes (ie~ geological studiesr mass balancesr analysis of Standard Reference Materials) or in response to a regulation that requires total sample decomposition

Even though the ICP-MS preparation method does not yield total results it still may conservatively measure the amount of metals that are environmentally available In the Research Project Summary Ambient Levels of Metals in New Jersey Soils prepared by the New Jersey Department of Environmental Protection (NJDEP 2003) where samples were prepared using SW846-3050 procedures it states

Metal concentrations reported in this summary were acid-extractable metals not necessarily total metal concentrations The extraction method (USEPA Method 3050) is a vigorous extraction method designed to remove all metals that could become environmentally available In practice the extraction method likely overestimates metals that could become available since the sample is refluxed with both concentrated nitric acid and hydrogen peroxide However by design the method will not extract chemicals incorporated in silica minerals as they are usually not mobile in the environment Thus concentrations reported in these studies may be lower than those indicated from analyses using methods designed to measure total metal concentrations such as x-ray fluorescence methods

Although references to uranium were not located it is expected that uranium will have some of the same type of mobility characteristics as other metals in the soil matrix The paper Ground Water Issue Behavior of Metals in Soil published by EPAs Superfund Technology Support Center for Ground Water (EPA 1992) describes how metals are distributed within soil The paper describes how metals are partitioned in various pools and how a rigorous extraction using HF is required to release metals present in the structure of primary or secondary minerals It is

3 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

expected that native metals (ie background concentrations) can be present in these structures The paper indicates that for four metals examined (cadmium copper zinc and nickel) the amount of metals that was bound in the mineral structure ranges from approximately 15-60 It would be expected that naturally occurring uranium would have similar characteristics

The National Institutes of Standards amp Technology (NIST) recognizes the differences between total digestion and method SW846-3050 and has issued Addendum to Certificates for the Certificate of Analysis (NIST 2003) for some Standard Reference Materials (SRMs) The Addendum to Certificates for SRM 2711 a Montana agricultural soil indicates that trace metals were measured at approximately 40 to 90 of their certified value when analyzed using SW846-3050 as the preparation method Even though uranium is not listed as a target analyte it is reasonable to expect that is would behave in a similar fashion

Sample Quanitation

The preparation method for alpha spec analysis includes addition of a tracer (U-232) to sample and is carried through the entire analytical scheme The amount of tracer is measured during analysis and the results for target compounds (the other uranium isotopes) are corrected for the amount of tracer recovery (yield) For example if amount of tracer measured is 50 of what was initially added to the sample the sample results are all corrected by dividing by 050 In terms of method quality assurancequality control (QAQC) the PRP QAPP indicates that tracer recovery in the range of 15-125 is considered acceptable and the analytical system is considered in control In other words up to 85 of the compound can be lost during sample preparation and analysis and the data is considered meeting all QAQC objectives This method of analysis is the equivalent to the technique of isotope dilution which is used for some non-radiological environmental analyses (eg EPA Method 1625 Semivolatile Organic Compounds by Isotope Dilution GCMS) Analyses using this technique although considered the most accurate are not widely employed due to the complexity and expense of using the isotope dilution technique

ICP-MS analysis measures the concentration of analyte with no correction for potential losses However QAQC procedures are in place to make sure the losses are not considered significant In addition to laboratory control samples matrix spike analysis measures the recovery for the target analyte that was over-spiked into the sample The PRP QAPP states that matrix spike recovery should be within the range of 75-125 Therefore up to 25 of the analyte could be lost and the method has still met the QAQC objectives If recovery is less than this then validation procedures acknowledge that the data may be biased low and sample results are flagged as estimated

Therefore because the alpha spec data is corrected for analyte loss during the analytical process while ICP-MS data is not some amount of positive bias is expected to be present in alpha spec data when compared to ICP-MS data

Considerations

In evaluating which analysis is appropriate for samples from the project neither method is inherently better than the other The following factors should be considered

1 What is being measured ICP-MS measures mass directly and data can be mathematically converted to activity Alpha spec measures activity directly and can be mathematically converted to mass

4 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

2 What are the objectives of the data in terms of measuring total versus environmentally available The ICP-MS method specified in the PRP QAPP measures uranium concentrations in the same manner that EPA has measured other metals concentrations Should uranium mass be measured in a typical way that will be used for measuring other metals of concern at the site (eg copper beryllium arsenic etc)

3 If ICP-MS is more cost effective that alpha spec should the option of using preparation method SW846-3052 be considered This should yield total uranium data similar to alpha spec analysis while still using the efficiencies of ICP-MS analysis

4 What units will the cleanup level for the site be in If its in mass units then ICPshyMS analysis may be more appropriate If its in activity units then alpha spec analysis may be warranted

References

Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts Technical Memorandum from Heath Downey to Bruce Thompson de maxim is inc September 26 2005 (MACTEC 2005)

Test Methods for Evaluating Solid Waste PhysicalChemical Methods SW-846 Third Edition with updates EPA 1996

Ambient Levels of Metals in New Jersey Soils Environmental Assessment and Risk Analysis Element Research Project Summary Division of Science Research and Technology State of New Jersey Department of Environmental Protection May2003

Ground Water Issue Behavior of Metals in Soil EPA Superfund Technology Support Center for Ground Water EPA540S-92018 October 1993

Certificate of Analysis Standard Reference Material 2711 Montana Soil National Institute of Standards amp Technology July 18 2003

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10000

MampE SoilSediment Results Total Uranium mgKg

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en aE

I I bull Total a u

I -

bull bull 100 1shy -middot- -middot bullbull bull

Ybull bull

10 100 1000 10000

Alpha Spec

Attachm

ent 1

10 -

100000

TOTAL URANIUM BY ICP-MS mgKg

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I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 17: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Figure 2

Alp

ha S

pect

osco

py (m

gkg

)

Fall 2004 Total Uranium No Super-Depleted

1000

100

10

1

y = 10604x R2 = 09049

1 10 100 1000

ICPMS (mgkg)

Fall 2004 Super-Depleted Linear Regression

Attachment 1Connolly Jayme

From Taylor Mel issagepam ail epa gov Sent Wednesday December 21 2005 932AM To brucetdemaximiscom jhuntdemaximiscom Subject Fw Uranium Analysis Issues

Rad-versus-ICPMS RADvsiCPMS_soilse MEvsPRP_ICPMSP memodoc (158 dPDF (29 KB) OF (25 KB)

brucejohn here is mampes take on the rns vs alpha spec analysis discrepancy

Forwarded by MelissaG TaylorR1USEPAUS on 12212005 0930 AM

Conroy Ed ltEdConroyregrn-ea ecomcomgt To

MelissaG TaylorR1USEPAUSregEPA 12152005 0416 cc PM

Subject FW Uranium Analysis Issues

Melissa

Heres Andys take on the alpha spec vs ICPMS debate along with a couple of graphs showing correlation between the two methods and MampE split data vs PRP data

Please call me on the cell phone tomorrow to discuss

Ed

Edward A Conroy PE

Project Manager

METCALF amp EDDYJAECOM

7812246880 ph

6172832942 cell

1

ed conroym-e aecom com Attachment 1

-----Original Message----shyFrom Schkuta Andrew Sent Thursday December 15 2005 358 PM To Conroy Ed Subject Uranium Analysis Issues

Ed

Please find attached a memorandum that summarizes my review of why ICP-MS and alpha spec data may not necessarily agree

ltltRad-versus-ICPMS memodocgtgt

In addition I have attached 2 preliminary charts that have the following draft information

ltltRADvsiCPMS soilsedPDFgtgt

The RADvsiCPMS soilsed chart is a draft plot of MampEs alpha spec versus ICP-MS results when data was available for both samples It shows the same general trend as the PRP data with fairly good agreement between the two analyses There appears to be a slight positive bias for alpha spec data at low concentrations but this is expected as discussed in my memo

ltltMEvsPRP ICPMSPDFgtgt

The MEvsPRP ICPMS is a draft plot of MampE vs the PRPs total uranium results as measured by ICP-MS -With the exception of a few outliers which will be further investigated there appears to be good agreement between MampEs and the PRPs data As a note the median percent difference between MampEs and the PRPs data was 1

Andrew Schkuta CHMM

Technical Specialist - Chemistry

METCALF amp EDDYJAECOM

701 Edgewater Drive

Wakefield MA 01880

(781) 224-6353

(781) 245-6293 fax

2

(See attached file Rad-versus-ICPMS Attachment 1memodoc) (See attached file RADvsiCPMS soilsedPDF) (See attached file MEvsPRP ICPMSPDF)

3

Attachment 1

Metcalf amp Eddy Inc

METCALFampEDDY

Memorandum

Date December 15 2005

To Ed Conroy

From Andrew Schkuta

Subject Uranium Analysis comparison of alpha spectrometry and ICP-MS results

File WA151 Nuclear Metals Distribution

Summary This memorandum discusses the differences between uranium analysis conducted by two different methods alpha spectroscopy (alpha spec) and inductively coupled plasma shymass spectrometry (ICP-MS) In summary the alpha spec data is expected to be biased high compared to the ICP-MS data but this is not indicative that either method is being performed incorrectly

In the September 25 2005 technical memorandum Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts (MACTEC 2005) data was presented comparing alpha spec and ICP-MS analysis for uranium analysis and it was proposed that future uranium analysis be performed by ICP-MS only For samples analyzed that were collected through Spring 2005 all samples were analyzed by ICP-MS with a subset (20) analyzed by alpha spec It was noted that there appeared to be a positive bias in the alpha spec data compared to the ICP-MS data and some concern has been voiced if analysis by ICP-MS only would meet project objectives This memorandum provides background on why the two analytical methods may yield differing results and discusses how they may impact project quality objectives

There are two primary reasons why alpha spec is expected to yield results that may be positively biased

1 Sample Preparation The sample preparation method for alpha spec is more rigorous The preparation method in alpha spec breaks down the silica mineral structure of the sample By design the ICP-MS sample preparation does not break down the mineral structure and is intended to remove all metals that could become environmentally available

2 Sample Quanitation The results for alpha spec are adjusted for analyte loss by correcting the result based on the amount of tracer recovered (yield) ICP-MS data is absolute ie data is not corrected for loss during sample preparation

2 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

These two reasons are discussed in further detail as follows

Sample Preparation

Radionuclide analysis traditionally uses a sample preparation method that includes hydrofluoric acid (HF) to break down the sample so that all of the uranium present in the sample is put into solution On the other hand standard EPA methods for metals analysis including ICP-MS do not measure the total metals present in a sample For example method SW846-3050 (EPA 1996) (as specified by the PRP in their approved QAPP for metals analysis by ICP-MS) states

This method is not a total digestion technique for most samples It is a very strong acid digestion that will dissolve almost all elements that could become environmentally available By design elements bound in silicate structures are not normally dissolved by this procedure as they are not usually mobile in the environment If absolute total digestion is required use Method 3052

As noted by the method a SW846 method is available to do a total digestion (equivalent to the alpha spec method) as method 3052 However method 3052 (EPA 1996) states (emphasis added)

Note This technique is not appropriate for regulatory applications that require the use of leachate preparations (ie Method 3050 Method 3051 Method 1311 Method 1312 Method 1310 Method 1320 Method 1330 Method 3031 Method 3040) This method is appropriate for those applications requiring a total decomposition for research purposes (ie~ geological studiesr mass balancesr analysis of Standard Reference Materials) or in response to a regulation that requires total sample decomposition

Even though the ICP-MS preparation method does not yield total results it still may conservatively measure the amount of metals that are environmentally available In the Research Project Summary Ambient Levels of Metals in New Jersey Soils prepared by the New Jersey Department of Environmental Protection (NJDEP 2003) where samples were prepared using SW846-3050 procedures it states

Metal concentrations reported in this summary were acid-extractable metals not necessarily total metal concentrations The extraction method (USEPA Method 3050) is a vigorous extraction method designed to remove all metals that could become environmentally available In practice the extraction method likely overestimates metals that could become available since the sample is refluxed with both concentrated nitric acid and hydrogen peroxide However by design the method will not extract chemicals incorporated in silica minerals as they are usually not mobile in the environment Thus concentrations reported in these studies may be lower than those indicated from analyses using methods designed to measure total metal concentrations such as x-ray fluorescence methods

Although references to uranium were not located it is expected that uranium will have some of the same type of mobility characteristics as other metals in the soil matrix The paper Ground Water Issue Behavior of Metals in Soil published by EPAs Superfund Technology Support Center for Ground Water (EPA 1992) describes how metals are distributed within soil The paper describes how metals are partitioned in various pools and how a rigorous extraction using HF is required to release metals present in the structure of primary or secondary minerals It is

3 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

expected that native metals (ie background concentrations) can be present in these structures The paper indicates that for four metals examined (cadmium copper zinc and nickel) the amount of metals that was bound in the mineral structure ranges from approximately 15-60 It would be expected that naturally occurring uranium would have similar characteristics

The National Institutes of Standards amp Technology (NIST) recognizes the differences between total digestion and method SW846-3050 and has issued Addendum to Certificates for the Certificate of Analysis (NIST 2003) for some Standard Reference Materials (SRMs) The Addendum to Certificates for SRM 2711 a Montana agricultural soil indicates that trace metals were measured at approximately 40 to 90 of their certified value when analyzed using SW846-3050 as the preparation method Even though uranium is not listed as a target analyte it is reasonable to expect that is would behave in a similar fashion

Sample Quanitation

The preparation method for alpha spec analysis includes addition of a tracer (U-232) to sample and is carried through the entire analytical scheme The amount of tracer is measured during analysis and the results for target compounds (the other uranium isotopes) are corrected for the amount of tracer recovery (yield) For example if amount of tracer measured is 50 of what was initially added to the sample the sample results are all corrected by dividing by 050 In terms of method quality assurancequality control (QAQC) the PRP QAPP indicates that tracer recovery in the range of 15-125 is considered acceptable and the analytical system is considered in control In other words up to 85 of the compound can be lost during sample preparation and analysis and the data is considered meeting all QAQC objectives This method of analysis is the equivalent to the technique of isotope dilution which is used for some non-radiological environmental analyses (eg EPA Method 1625 Semivolatile Organic Compounds by Isotope Dilution GCMS) Analyses using this technique although considered the most accurate are not widely employed due to the complexity and expense of using the isotope dilution technique

ICP-MS analysis measures the concentration of analyte with no correction for potential losses However QAQC procedures are in place to make sure the losses are not considered significant In addition to laboratory control samples matrix spike analysis measures the recovery for the target analyte that was over-spiked into the sample The PRP QAPP states that matrix spike recovery should be within the range of 75-125 Therefore up to 25 of the analyte could be lost and the method has still met the QAQC objectives If recovery is less than this then validation procedures acknowledge that the data may be biased low and sample results are flagged as estimated

Therefore because the alpha spec data is corrected for analyte loss during the analytical process while ICP-MS data is not some amount of positive bias is expected to be present in alpha spec data when compared to ICP-MS data

Considerations

In evaluating which analysis is appropriate for samples from the project neither method is inherently better than the other The following factors should be considered

1 What is being measured ICP-MS measures mass directly and data can be mathematically converted to activity Alpha spec measures activity directly and can be mathematically converted to mass

4 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

2 What are the objectives of the data in terms of measuring total versus environmentally available The ICP-MS method specified in the PRP QAPP measures uranium concentrations in the same manner that EPA has measured other metals concentrations Should uranium mass be measured in a typical way that will be used for measuring other metals of concern at the site (eg copper beryllium arsenic etc)

3 If ICP-MS is more cost effective that alpha spec should the option of using preparation method SW846-3052 be considered This should yield total uranium data similar to alpha spec analysis while still using the efficiencies of ICP-MS analysis

4 What units will the cleanup level for the site be in If its in mass units then ICPshyMS analysis may be more appropriate If its in activity units then alpha spec analysis may be warranted

References

Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts Technical Memorandum from Heath Downey to Bruce Thompson de maxim is inc September 26 2005 (MACTEC 2005)

Test Methods for Evaluating Solid Waste PhysicalChemical Methods SW-846 Third Edition with updates EPA 1996

Ambient Levels of Metals in New Jersey Soils Environmental Assessment and Risk Analysis Element Research Project Summary Division of Science Research and Technology State of New Jersey Department of Environmental Protection May2003

Ground Water Issue Behavior of Metals in Soil EPA Superfund Technology Support Center for Ground Water EPA540S-92018 October 1993

Certificate of Analysis Standard Reference Material 2711 Montana Soil National Institute of Standards amp Technology July 18 2003

bull bull

bull bull bullbull bull

10000

MampE SoilSediment Results Total Uranium mgKg

bull

bull - ~-~-----~~---~~-~--~~1000

en aE

I I bull Total a u

I -

bull bull 100 1shy -middot- -middot bullbull bull

Ybull bull

10 100 1000 10000

Alpha Spec

Attachm

ent 1

10 -

100000

TOTAL URANIUM BY ICP-MS mgKg

bull

bull 10000

I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 18: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 1Connolly Jayme

From Taylor Mel issagepam ail epa gov Sent Wednesday December 21 2005 932AM To brucetdemaximiscom jhuntdemaximiscom Subject Fw Uranium Analysis Issues

Rad-versus-ICPMS RADvsiCPMS_soilse MEvsPRP_ICPMSP memodoc (158 dPDF (29 KB) OF (25 KB)

brucejohn here is mampes take on the rns vs alpha spec analysis discrepancy

Forwarded by MelissaG TaylorR1USEPAUS on 12212005 0930 AM

Conroy Ed ltEdConroyregrn-ea ecomcomgt To

MelissaG TaylorR1USEPAUSregEPA 12152005 0416 cc PM

Subject FW Uranium Analysis Issues

Melissa

Heres Andys take on the alpha spec vs ICPMS debate along with a couple of graphs showing correlation between the two methods and MampE split data vs PRP data

Please call me on the cell phone tomorrow to discuss

Ed

Edward A Conroy PE

Project Manager

METCALF amp EDDYJAECOM

7812246880 ph

6172832942 cell

1

ed conroym-e aecom com Attachment 1

-----Original Message----shyFrom Schkuta Andrew Sent Thursday December 15 2005 358 PM To Conroy Ed Subject Uranium Analysis Issues

Ed

Please find attached a memorandum that summarizes my review of why ICP-MS and alpha spec data may not necessarily agree

ltltRad-versus-ICPMS memodocgtgt

In addition I have attached 2 preliminary charts that have the following draft information

ltltRADvsiCPMS soilsedPDFgtgt

The RADvsiCPMS soilsed chart is a draft plot of MampEs alpha spec versus ICP-MS results when data was available for both samples It shows the same general trend as the PRP data with fairly good agreement between the two analyses There appears to be a slight positive bias for alpha spec data at low concentrations but this is expected as discussed in my memo

ltltMEvsPRP ICPMSPDFgtgt

The MEvsPRP ICPMS is a draft plot of MampE vs the PRPs total uranium results as measured by ICP-MS -With the exception of a few outliers which will be further investigated there appears to be good agreement between MampEs and the PRPs data As a note the median percent difference between MampEs and the PRPs data was 1

Andrew Schkuta CHMM

Technical Specialist - Chemistry

METCALF amp EDDYJAECOM

701 Edgewater Drive

Wakefield MA 01880

(781) 224-6353

(781) 245-6293 fax

2

(See attached file Rad-versus-ICPMS Attachment 1memodoc) (See attached file RADvsiCPMS soilsedPDF) (See attached file MEvsPRP ICPMSPDF)

3

Attachment 1

Metcalf amp Eddy Inc

METCALFampEDDY

Memorandum

Date December 15 2005

To Ed Conroy

From Andrew Schkuta

Subject Uranium Analysis comparison of alpha spectrometry and ICP-MS results

File WA151 Nuclear Metals Distribution

Summary This memorandum discusses the differences between uranium analysis conducted by two different methods alpha spectroscopy (alpha spec) and inductively coupled plasma shymass spectrometry (ICP-MS) In summary the alpha spec data is expected to be biased high compared to the ICP-MS data but this is not indicative that either method is being performed incorrectly

In the September 25 2005 technical memorandum Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts (MACTEC 2005) data was presented comparing alpha spec and ICP-MS analysis for uranium analysis and it was proposed that future uranium analysis be performed by ICP-MS only For samples analyzed that were collected through Spring 2005 all samples were analyzed by ICP-MS with a subset (20) analyzed by alpha spec It was noted that there appeared to be a positive bias in the alpha spec data compared to the ICP-MS data and some concern has been voiced if analysis by ICP-MS only would meet project objectives This memorandum provides background on why the two analytical methods may yield differing results and discusses how they may impact project quality objectives

There are two primary reasons why alpha spec is expected to yield results that may be positively biased

1 Sample Preparation The sample preparation method for alpha spec is more rigorous The preparation method in alpha spec breaks down the silica mineral structure of the sample By design the ICP-MS sample preparation does not break down the mineral structure and is intended to remove all metals that could become environmentally available

2 Sample Quanitation The results for alpha spec are adjusted for analyte loss by correcting the result based on the amount of tracer recovered (yield) ICP-MS data is absolute ie data is not corrected for loss during sample preparation

2 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

These two reasons are discussed in further detail as follows

Sample Preparation

Radionuclide analysis traditionally uses a sample preparation method that includes hydrofluoric acid (HF) to break down the sample so that all of the uranium present in the sample is put into solution On the other hand standard EPA methods for metals analysis including ICP-MS do not measure the total metals present in a sample For example method SW846-3050 (EPA 1996) (as specified by the PRP in their approved QAPP for metals analysis by ICP-MS) states

This method is not a total digestion technique for most samples It is a very strong acid digestion that will dissolve almost all elements that could become environmentally available By design elements bound in silicate structures are not normally dissolved by this procedure as they are not usually mobile in the environment If absolute total digestion is required use Method 3052

As noted by the method a SW846 method is available to do a total digestion (equivalent to the alpha spec method) as method 3052 However method 3052 (EPA 1996) states (emphasis added)

Note This technique is not appropriate for regulatory applications that require the use of leachate preparations (ie Method 3050 Method 3051 Method 1311 Method 1312 Method 1310 Method 1320 Method 1330 Method 3031 Method 3040) This method is appropriate for those applications requiring a total decomposition for research purposes (ie~ geological studiesr mass balancesr analysis of Standard Reference Materials) or in response to a regulation that requires total sample decomposition

Even though the ICP-MS preparation method does not yield total results it still may conservatively measure the amount of metals that are environmentally available In the Research Project Summary Ambient Levels of Metals in New Jersey Soils prepared by the New Jersey Department of Environmental Protection (NJDEP 2003) where samples were prepared using SW846-3050 procedures it states

Metal concentrations reported in this summary were acid-extractable metals not necessarily total metal concentrations The extraction method (USEPA Method 3050) is a vigorous extraction method designed to remove all metals that could become environmentally available In practice the extraction method likely overestimates metals that could become available since the sample is refluxed with both concentrated nitric acid and hydrogen peroxide However by design the method will not extract chemicals incorporated in silica minerals as they are usually not mobile in the environment Thus concentrations reported in these studies may be lower than those indicated from analyses using methods designed to measure total metal concentrations such as x-ray fluorescence methods

Although references to uranium were not located it is expected that uranium will have some of the same type of mobility characteristics as other metals in the soil matrix The paper Ground Water Issue Behavior of Metals in Soil published by EPAs Superfund Technology Support Center for Ground Water (EPA 1992) describes how metals are distributed within soil The paper describes how metals are partitioned in various pools and how a rigorous extraction using HF is required to release metals present in the structure of primary or secondary minerals It is

3 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

expected that native metals (ie background concentrations) can be present in these structures The paper indicates that for four metals examined (cadmium copper zinc and nickel) the amount of metals that was bound in the mineral structure ranges from approximately 15-60 It would be expected that naturally occurring uranium would have similar characteristics

The National Institutes of Standards amp Technology (NIST) recognizes the differences between total digestion and method SW846-3050 and has issued Addendum to Certificates for the Certificate of Analysis (NIST 2003) for some Standard Reference Materials (SRMs) The Addendum to Certificates for SRM 2711 a Montana agricultural soil indicates that trace metals were measured at approximately 40 to 90 of their certified value when analyzed using SW846-3050 as the preparation method Even though uranium is not listed as a target analyte it is reasonable to expect that is would behave in a similar fashion

Sample Quanitation

The preparation method for alpha spec analysis includes addition of a tracer (U-232) to sample and is carried through the entire analytical scheme The amount of tracer is measured during analysis and the results for target compounds (the other uranium isotopes) are corrected for the amount of tracer recovery (yield) For example if amount of tracer measured is 50 of what was initially added to the sample the sample results are all corrected by dividing by 050 In terms of method quality assurancequality control (QAQC) the PRP QAPP indicates that tracer recovery in the range of 15-125 is considered acceptable and the analytical system is considered in control In other words up to 85 of the compound can be lost during sample preparation and analysis and the data is considered meeting all QAQC objectives This method of analysis is the equivalent to the technique of isotope dilution which is used for some non-radiological environmental analyses (eg EPA Method 1625 Semivolatile Organic Compounds by Isotope Dilution GCMS) Analyses using this technique although considered the most accurate are not widely employed due to the complexity and expense of using the isotope dilution technique

ICP-MS analysis measures the concentration of analyte with no correction for potential losses However QAQC procedures are in place to make sure the losses are not considered significant In addition to laboratory control samples matrix spike analysis measures the recovery for the target analyte that was over-spiked into the sample The PRP QAPP states that matrix spike recovery should be within the range of 75-125 Therefore up to 25 of the analyte could be lost and the method has still met the QAQC objectives If recovery is less than this then validation procedures acknowledge that the data may be biased low and sample results are flagged as estimated

Therefore because the alpha spec data is corrected for analyte loss during the analytical process while ICP-MS data is not some amount of positive bias is expected to be present in alpha spec data when compared to ICP-MS data

Considerations

In evaluating which analysis is appropriate for samples from the project neither method is inherently better than the other The following factors should be considered

1 What is being measured ICP-MS measures mass directly and data can be mathematically converted to activity Alpha spec measures activity directly and can be mathematically converted to mass

4 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

2 What are the objectives of the data in terms of measuring total versus environmentally available The ICP-MS method specified in the PRP QAPP measures uranium concentrations in the same manner that EPA has measured other metals concentrations Should uranium mass be measured in a typical way that will be used for measuring other metals of concern at the site (eg copper beryllium arsenic etc)

3 If ICP-MS is more cost effective that alpha spec should the option of using preparation method SW846-3052 be considered This should yield total uranium data similar to alpha spec analysis while still using the efficiencies of ICP-MS analysis

4 What units will the cleanup level for the site be in If its in mass units then ICPshyMS analysis may be more appropriate If its in activity units then alpha spec analysis may be warranted

References

Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts Technical Memorandum from Heath Downey to Bruce Thompson de maxim is inc September 26 2005 (MACTEC 2005)

Test Methods for Evaluating Solid Waste PhysicalChemical Methods SW-846 Third Edition with updates EPA 1996

Ambient Levels of Metals in New Jersey Soils Environmental Assessment and Risk Analysis Element Research Project Summary Division of Science Research and Technology State of New Jersey Department of Environmental Protection May2003

Ground Water Issue Behavior of Metals in Soil EPA Superfund Technology Support Center for Ground Water EPA540S-92018 October 1993

Certificate of Analysis Standard Reference Material 2711 Montana Soil National Institute of Standards amp Technology July 18 2003

bull bull

bull bull bullbull bull

10000

MampE SoilSediment Results Total Uranium mgKg

bull

bull - ~-~-----~~---~~-~--~~1000

en aE

I I bull Total a u

I -

bull bull 100 1shy -middot- -middot bullbull bull

Ybull bull

10 100 1000 10000

Alpha Spec

Attachm

ent 1

10 -

100000

TOTAL URANIUM BY ICP-MS mgKg

bull

bull 10000

I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 19: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

ed conroym-e aecom com Attachment 1

-----Original Message----shyFrom Schkuta Andrew Sent Thursday December 15 2005 358 PM To Conroy Ed Subject Uranium Analysis Issues

Ed

Please find attached a memorandum that summarizes my review of why ICP-MS and alpha spec data may not necessarily agree

ltltRad-versus-ICPMS memodocgtgt

In addition I have attached 2 preliminary charts that have the following draft information

ltltRADvsiCPMS soilsedPDFgtgt

The RADvsiCPMS soilsed chart is a draft plot of MampEs alpha spec versus ICP-MS results when data was available for both samples It shows the same general trend as the PRP data with fairly good agreement between the two analyses There appears to be a slight positive bias for alpha spec data at low concentrations but this is expected as discussed in my memo

ltltMEvsPRP ICPMSPDFgtgt

The MEvsPRP ICPMS is a draft plot of MampE vs the PRPs total uranium results as measured by ICP-MS -With the exception of a few outliers which will be further investigated there appears to be good agreement between MampEs and the PRPs data As a note the median percent difference between MampEs and the PRPs data was 1

Andrew Schkuta CHMM

Technical Specialist - Chemistry

METCALF amp EDDYJAECOM

701 Edgewater Drive

Wakefield MA 01880

(781) 224-6353

(781) 245-6293 fax

2

(See attached file Rad-versus-ICPMS Attachment 1memodoc) (See attached file RADvsiCPMS soilsedPDF) (See attached file MEvsPRP ICPMSPDF)

3

Attachment 1

Metcalf amp Eddy Inc

METCALFampEDDY

Memorandum

Date December 15 2005

To Ed Conroy

From Andrew Schkuta

Subject Uranium Analysis comparison of alpha spectrometry and ICP-MS results

File WA151 Nuclear Metals Distribution

Summary This memorandum discusses the differences between uranium analysis conducted by two different methods alpha spectroscopy (alpha spec) and inductively coupled plasma shymass spectrometry (ICP-MS) In summary the alpha spec data is expected to be biased high compared to the ICP-MS data but this is not indicative that either method is being performed incorrectly

In the September 25 2005 technical memorandum Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts (MACTEC 2005) data was presented comparing alpha spec and ICP-MS analysis for uranium analysis and it was proposed that future uranium analysis be performed by ICP-MS only For samples analyzed that were collected through Spring 2005 all samples were analyzed by ICP-MS with a subset (20) analyzed by alpha spec It was noted that there appeared to be a positive bias in the alpha spec data compared to the ICP-MS data and some concern has been voiced if analysis by ICP-MS only would meet project objectives This memorandum provides background on why the two analytical methods may yield differing results and discusses how they may impact project quality objectives

There are two primary reasons why alpha spec is expected to yield results that may be positively biased

1 Sample Preparation The sample preparation method for alpha spec is more rigorous The preparation method in alpha spec breaks down the silica mineral structure of the sample By design the ICP-MS sample preparation does not break down the mineral structure and is intended to remove all metals that could become environmentally available

2 Sample Quanitation The results for alpha spec are adjusted for analyte loss by correcting the result based on the amount of tracer recovered (yield) ICP-MS data is absolute ie data is not corrected for loss during sample preparation

2 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

These two reasons are discussed in further detail as follows

Sample Preparation

Radionuclide analysis traditionally uses a sample preparation method that includes hydrofluoric acid (HF) to break down the sample so that all of the uranium present in the sample is put into solution On the other hand standard EPA methods for metals analysis including ICP-MS do not measure the total metals present in a sample For example method SW846-3050 (EPA 1996) (as specified by the PRP in their approved QAPP for metals analysis by ICP-MS) states

This method is not a total digestion technique for most samples It is a very strong acid digestion that will dissolve almost all elements that could become environmentally available By design elements bound in silicate structures are not normally dissolved by this procedure as they are not usually mobile in the environment If absolute total digestion is required use Method 3052

As noted by the method a SW846 method is available to do a total digestion (equivalent to the alpha spec method) as method 3052 However method 3052 (EPA 1996) states (emphasis added)

Note This technique is not appropriate for regulatory applications that require the use of leachate preparations (ie Method 3050 Method 3051 Method 1311 Method 1312 Method 1310 Method 1320 Method 1330 Method 3031 Method 3040) This method is appropriate for those applications requiring a total decomposition for research purposes (ie~ geological studiesr mass balancesr analysis of Standard Reference Materials) or in response to a regulation that requires total sample decomposition

Even though the ICP-MS preparation method does not yield total results it still may conservatively measure the amount of metals that are environmentally available In the Research Project Summary Ambient Levels of Metals in New Jersey Soils prepared by the New Jersey Department of Environmental Protection (NJDEP 2003) where samples were prepared using SW846-3050 procedures it states

Metal concentrations reported in this summary were acid-extractable metals not necessarily total metal concentrations The extraction method (USEPA Method 3050) is a vigorous extraction method designed to remove all metals that could become environmentally available In practice the extraction method likely overestimates metals that could become available since the sample is refluxed with both concentrated nitric acid and hydrogen peroxide However by design the method will not extract chemicals incorporated in silica minerals as they are usually not mobile in the environment Thus concentrations reported in these studies may be lower than those indicated from analyses using methods designed to measure total metal concentrations such as x-ray fluorescence methods

Although references to uranium were not located it is expected that uranium will have some of the same type of mobility characteristics as other metals in the soil matrix The paper Ground Water Issue Behavior of Metals in Soil published by EPAs Superfund Technology Support Center for Ground Water (EPA 1992) describes how metals are distributed within soil The paper describes how metals are partitioned in various pools and how a rigorous extraction using HF is required to release metals present in the structure of primary or secondary minerals It is

3 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

expected that native metals (ie background concentrations) can be present in these structures The paper indicates that for four metals examined (cadmium copper zinc and nickel) the amount of metals that was bound in the mineral structure ranges from approximately 15-60 It would be expected that naturally occurring uranium would have similar characteristics

The National Institutes of Standards amp Technology (NIST) recognizes the differences between total digestion and method SW846-3050 and has issued Addendum to Certificates for the Certificate of Analysis (NIST 2003) for some Standard Reference Materials (SRMs) The Addendum to Certificates for SRM 2711 a Montana agricultural soil indicates that trace metals were measured at approximately 40 to 90 of their certified value when analyzed using SW846-3050 as the preparation method Even though uranium is not listed as a target analyte it is reasonable to expect that is would behave in a similar fashion

Sample Quanitation

The preparation method for alpha spec analysis includes addition of a tracer (U-232) to sample and is carried through the entire analytical scheme The amount of tracer is measured during analysis and the results for target compounds (the other uranium isotopes) are corrected for the amount of tracer recovery (yield) For example if amount of tracer measured is 50 of what was initially added to the sample the sample results are all corrected by dividing by 050 In terms of method quality assurancequality control (QAQC) the PRP QAPP indicates that tracer recovery in the range of 15-125 is considered acceptable and the analytical system is considered in control In other words up to 85 of the compound can be lost during sample preparation and analysis and the data is considered meeting all QAQC objectives This method of analysis is the equivalent to the technique of isotope dilution which is used for some non-radiological environmental analyses (eg EPA Method 1625 Semivolatile Organic Compounds by Isotope Dilution GCMS) Analyses using this technique although considered the most accurate are not widely employed due to the complexity and expense of using the isotope dilution technique

ICP-MS analysis measures the concentration of analyte with no correction for potential losses However QAQC procedures are in place to make sure the losses are not considered significant In addition to laboratory control samples matrix spike analysis measures the recovery for the target analyte that was over-spiked into the sample The PRP QAPP states that matrix spike recovery should be within the range of 75-125 Therefore up to 25 of the analyte could be lost and the method has still met the QAQC objectives If recovery is less than this then validation procedures acknowledge that the data may be biased low and sample results are flagged as estimated

Therefore because the alpha spec data is corrected for analyte loss during the analytical process while ICP-MS data is not some amount of positive bias is expected to be present in alpha spec data when compared to ICP-MS data

Considerations

In evaluating which analysis is appropriate for samples from the project neither method is inherently better than the other The following factors should be considered

1 What is being measured ICP-MS measures mass directly and data can be mathematically converted to activity Alpha spec measures activity directly and can be mathematically converted to mass

4 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

2 What are the objectives of the data in terms of measuring total versus environmentally available The ICP-MS method specified in the PRP QAPP measures uranium concentrations in the same manner that EPA has measured other metals concentrations Should uranium mass be measured in a typical way that will be used for measuring other metals of concern at the site (eg copper beryllium arsenic etc)

3 If ICP-MS is more cost effective that alpha spec should the option of using preparation method SW846-3052 be considered This should yield total uranium data similar to alpha spec analysis while still using the efficiencies of ICP-MS analysis

4 What units will the cleanup level for the site be in If its in mass units then ICPshyMS analysis may be more appropriate If its in activity units then alpha spec analysis may be warranted

References

Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts Technical Memorandum from Heath Downey to Bruce Thompson de maxim is inc September 26 2005 (MACTEC 2005)

Test Methods for Evaluating Solid Waste PhysicalChemical Methods SW-846 Third Edition with updates EPA 1996

Ambient Levels of Metals in New Jersey Soils Environmental Assessment and Risk Analysis Element Research Project Summary Division of Science Research and Technology State of New Jersey Department of Environmental Protection May2003

Ground Water Issue Behavior of Metals in Soil EPA Superfund Technology Support Center for Ground Water EPA540S-92018 October 1993

Certificate of Analysis Standard Reference Material 2711 Montana Soil National Institute of Standards amp Technology July 18 2003

bull bull

bull bull bullbull bull

10000

MampE SoilSediment Results Total Uranium mgKg

bull

bull - ~-~-----~~---~~-~--~~1000

en aE

I I bull Total a u

I -

bull bull 100 1shy -middot- -middot bullbull bull

Ybull bull

10 100 1000 10000

Alpha Spec

Attachm

ent 1

10 -

100000

TOTAL URANIUM BY ICP-MS mgKg

bull

bull 10000

I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 20: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

(See attached file Rad-versus-ICPMS Attachment 1memodoc) (See attached file RADvsiCPMS soilsedPDF) (See attached file MEvsPRP ICPMSPDF)

3

Attachment 1

Metcalf amp Eddy Inc

METCALFampEDDY

Memorandum

Date December 15 2005

To Ed Conroy

From Andrew Schkuta

Subject Uranium Analysis comparison of alpha spectrometry and ICP-MS results

File WA151 Nuclear Metals Distribution

Summary This memorandum discusses the differences between uranium analysis conducted by two different methods alpha spectroscopy (alpha spec) and inductively coupled plasma shymass spectrometry (ICP-MS) In summary the alpha spec data is expected to be biased high compared to the ICP-MS data but this is not indicative that either method is being performed incorrectly

In the September 25 2005 technical memorandum Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts (MACTEC 2005) data was presented comparing alpha spec and ICP-MS analysis for uranium analysis and it was proposed that future uranium analysis be performed by ICP-MS only For samples analyzed that were collected through Spring 2005 all samples were analyzed by ICP-MS with a subset (20) analyzed by alpha spec It was noted that there appeared to be a positive bias in the alpha spec data compared to the ICP-MS data and some concern has been voiced if analysis by ICP-MS only would meet project objectives This memorandum provides background on why the two analytical methods may yield differing results and discusses how they may impact project quality objectives

There are two primary reasons why alpha spec is expected to yield results that may be positively biased

1 Sample Preparation The sample preparation method for alpha spec is more rigorous The preparation method in alpha spec breaks down the silica mineral structure of the sample By design the ICP-MS sample preparation does not break down the mineral structure and is intended to remove all metals that could become environmentally available

2 Sample Quanitation The results for alpha spec are adjusted for analyte loss by correcting the result based on the amount of tracer recovered (yield) ICP-MS data is absolute ie data is not corrected for loss during sample preparation

2 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

These two reasons are discussed in further detail as follows

Sample Preparation

Radionuclide analysis traditionally uses a sample preparation method that includes hydrofluoric acid (HF) to break down the sample so that all of the uranium present in the sample is put into solution On the other hand standard EPA methods for metals analysis including ICP-MS do not measure the total metals present in a sample For example method SW846-3050 (EPA 1996) (as specified by the PRP in their approved QAPP for metals analysis by ICP-MS) states

This method is not a total digestion technique for most samples It is a very strong acid digestion that will dissolve almost all elements that could become environmentally available By design elements bound in silicate structures are not normally dissolved by this procedure as they are not usually mobile in the environment If absolute total digestion is required use Method 3052

As noted by the method a SW846 method is available to do a total digestion (equivalent to the alpha spec method) as method 3052 However method 3052 (EPA 1996) states (emphasis added)

Note This technique is not appropriate for regulatory applications that require the use of leachate preparations (ie Method 3050 Method 3051 Method 1311 Method 1312 Method 1310 Method 1320 Method 1330 Method 3031 Method 3040) This method is appropriate for those applications requiring a total decomposition for research purposes (ie~ geological studiesr mass balancesr analysis of Standard Reference Materials) or in response to a regulation that requires total sample decomposition

Even though the ICP-MS preparation method does not yield total results it still may conservatively measure the amount of metals that are environmentally available In the Research Project Summary Ambient Levels of Metals in New Jersey Soils prepared by the New Jersey Department of Environmental Protection (NJDEP 2003) where samples were prepared using SW846-3050 procedures it states

Metal concentrations reported in this summary were acid-extractable metals not necessarily total metal concentrations The extraction method (USEPA Method 3050) is a vigorous extraction method designed to remove all metals that could become environmentally available In practice the extraction method likely overestimates metals that could become available since the sample is refluxed with both concentrated nitric acid and hydrogen peroxide However by design the method will not extract chemicals incorporated in silica minerals as they are usually not mobile in the environment Thus concentrations reported in these studies may be lower than those indicated from analyses using methods designed to measure total metal concentrations such as x-ray fluorescence methods

Although references to uranium were not located it is expected that uranium will have some of the same type of mobility characteristics as other metals in the soil matrix The paper Ground Water Issue Behavior of Metals in Soil published by EPAs Superfund Technology Support Center for Ground Water (EPA 1992) describes how metals are distributed within soil The paper describes how metals are partitioned in various pools and how a rigorous extraction using HF is required to release metals present in the structure of primary or secondary minerals It is

3 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

expected that native metals (ie background concentrations) can be present in these structures The paper indicates that for four metals examined (cadmium copper zinc and nickel) the amount of metals that was bound in the mineral structure ranges from approximately 15-60 It would be expected that naturally occurring uranium would have similar characteristics

The National Institutes of Standards amp Technology (NIST) recognizes the differences between total digestion and method SW846-3050 and has issued Addendum to Certificates for the Certificate of Analysis (NIST 2003) for some Standard Reference Materials (SRMs) The Addendum to Certificates for SRM 2711 a Montana agricultural soil indicates that trace metals were measured at approximately 40 to 90 of their certified value when analyzed using SW846-3050 as the preparation method Even though uranium is not listed as a target analyte it is reasonable to expect that is would behave in a similar fashion

Sample Quanitation

The preparation method for alpha spec analysis includes addition of a tracer (U-232) to sample and is carried through the entire analytical scheme The amount of tracer is measured during analysis and the results for target compounds (the other uranium isotopes) are corrected for the amount of tracer recovery (yield) For example if amount of tracer measured is 50 of what was initially added to the sample the sample results are all corrected by dividing by 050 In terms of method quality assurancequality control (QAQC) the PRP QAPP indicates that tracer recovery in the range of 15-125 is considered acceptable and the analytical system is considered in control In other words up to 85 of the compound can be lost during sample preparation and analysis and the data is considered meeting all QAQC objectives This method of analysis is the equivalent to the technique of isotope dilution which is used for some non-radiological environmental analyses (eg EPA Method 1625 Semivolatile Organic Compounds by Isotope Dilution GCMS) Analyses using this technique although considered the most accurate are not widely employed due to the complexity and expense of using the isotope dilution technique

ICP-MS analysis measures the concentration of analyte with no correction for potential losses However QAQC procedures are in place to make sure the losses are not considered significant In addition to laboratory control samples matrix spike analysis measures the recovery for the target analyte that was over-spiked into the sample The PRP QAPP states that matrix spike recovery should be within the range of 75-125 Therefore up to 25 of the analyte could be lost and the method has still met the QAQC objectives If recovery is less than this then validation procedures acknowledge that the data may be biased low and sample results are flagged as estimated

Therefore because the alpha spec data is corrected for analyte loss during the analytical process while ICP-MS data is not some amount of positive bias is expected to be present in alpha spec data when compared to ICP-MS data

Considerations

In evaluating which analysis is appropriate for samples from the project neither method is inherently better than the other The following factors should be considered

1 What is being measured ICP-MS measures mass directly and data can be mathematically converted to activity Alpha spec measures activity directly and can be mathematically converted to mass

4 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

2 What are the objectives of the data in terms of measuring total versus environmentally available The ICP-MS method specified in the PRP QAPP measures uranium concentrations in the same manner that EPA has measured other metals concentrations Should uranium mass be measured in a typical way that will be used for measuring other metals of concern at the site (eg copper beryllium arsenic etc)

3 If ICP-MS is more cost effective that alpha spec should the option of using preparation method SW846-3052 be considered This should yield total uranium data similar to alpha spec analysis while still using the efficiencies of ICP-MS analysis

4 What units will the cleanup level for the site be in If its in mass units then ICPshyMS analysis may be more appropriate If its in activity units then alpha spec analysis may be warranted

References

Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts Technical Memorandum from Heath Downey to Bruce Thompson de maxim is inc September 26 2005 (MACTEC 2005)

Test Methods for Evaluating Solid Waste PhysicalChemical Methods SW-846 Third Edition with updates EPA 1996

Ambient Levels of Metals in New Jersey Soils Environmental Assessment and Risk Analysis Element Research Project Summary Division of Science Research and Technology State of New Jersey Department of Environmental Protection May2003

Ground Water Issue Behavior of Metals in Soil EPA Superfund Technology Support Center for Ground Water EPA540S-92018 October 1993

Certificate of Analysis Standard Reference Material 2711 Montana Soil National Institute of Standards amp Technology July 18 2003

bull bull

bull bull bullbull bull

10000

MampE SoilSediment Results Total Uranium mgKg

bull

bull - ~-~-----~~---~~-~--~~1000

en aE

I I bull Total a u

I -

bull bull 100 1shy -middot- -middot bullbull bull

Ybull bull

10 100 1000 10000

Alpha Spec

Attachm

ent 1

10 -

100000

TOTAL URANIUM BY ICP-MS mgKg

bull

bull 10000

I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 21: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 1

Metcalf amp Eddy Inc

METCALFampEDDY

Memorandum

Date December 15 2005

To Ed Conroy

From Andrew Schkuta

Subject Uranium Analysis comparison of alpha spectrometry and ICP-MS results

File WA151 Nuclear Metals Distribution

Summary This memorandum discusses the differences between uranium analysis conducted by two different methods alpha spectroscopy (alpha spec) and inductively coupled plasma shymass spectrometry (ICP-MS) In summary the alpha spec data is expected to be biased high compared to the ICP-MS data but this is not indicative that either method is being performed incorrectly

In the September 25 2005 technical memorandum Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts (MACTEC 2005) data was presented comparing alpha spec and ICP-MS analysis for uranium analysis and it was proposed that future uranium analysis be performed by ICP-MS only For samples analyzed that were collected through Spring 2005 all samples were analyzed by ICP-MS with a subset (20) analyzed by alpha spec It was noted that there appeared to be a positive bias in the alpha spec data compared to the ICP-MS data and some concern has been voiced if analysis by ICP-MS only would meet project objectives This memorandum provides background on why the two analytical methods may yield differing results and discusses how they may impact project quality objectives

There are two primary reasons why alpha spec is expected to yield results that may be positively biased

1 Sample Preparation The sample preparation method for alpha spec is more rigorous The preparation method in alpha spec breaks down the silica mineral structure of the sample By design the ICP-MS sample preparation does not break down the mineral structure and is intended to remove all metals that could become environmentally available

2 Sample Quanitation The results for alpha spec are adjusted for analyte loss by correcting the result based on the amount of tracer recovered (yield) ICP-MS data is absolute ie data is not corrected for loss during sample preparation

2 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

These two reasons are discussed in further detail as follows

Sample Preparation

Radionuclide analysis traditionally uses a sample preparation method that includes hydrofluoric acid (HF) to break down the sample so that all of the uranium present in the sample is put into solution On the other hand standard EPA methods for metals analysis including ICP-MS do not measure the total metals present in a sample For example method SW846-3050 (EPA 1996) (as specified by the PRP in their approved QAPP for metals analysis by ICP-MS) states

This method is not a total digestion technique for most samples It is a very strong acid digestion that will dissolve almost all elements that could become environmentally available By design elements bound in silicate structures are not normally dissolved by this procedure as they are not usually mobile in the environment If absolute total digestion is required use Method 3052

As noted by the method a SW846 method is available to do a total digestion (equivalent to the alpha spec method) as method 3052 However method 3052 (EPA 1996) states (emphasis added)

Note This technique is not appropriate for regulatory applications that require the use of leachate preparations (ie Method 3050 Method 3051 Method 1311 Method 1312 Method 1310 Method 1320 Method 1330 Method 3031 Method 3040) This method is appropriate for those applications requiring a total decomposition for research purposes (ie~ geological studiesr mass balancesr analysis of Standard Reference Materials) or in response to a regulation that requires total sample decomposition

Even though the ICP-MS preparation method does not yield total results it still may conservatively measure the amount of metals that are environmentally available In the Research Project Summary Ambient Levels of Metals in New Jersey Soils prepared by the New Jersey Department of Environmental Protection (NJDEP 2003) where samples were prepared using SW846-3050 procedures it states

Metal concentrations reported in this summary were acid-extractable metals not necessarily total metal concentrations The extraction method (USEPA Method 3050) is a vigorous extraction method designed to remove all metals that could become environmentally available In practice the extraction method likely overestimates metals that could become available since the sample is refluxed with both concentrated nitric acid and hydrogen peroxide However by design the method will not extract chemicals incorporated in silica minerals as they are usually not mobile in the environment Thus concentrations reported in these studies may be lower than those indicated from analyses using methods designed to measure total metal concentrations such as x-ray fluorescence methods

Although references to uranium were not located it is expected that uranium will have some of the same type of mobility characteristics as other metals in the soil matrix The paper Ground Water Issue Behavior of Metals in Soil published by EPAs Superfund Technology Support Center for Ground Water (EPA 1992) describes how metals are distributed within soil The paper describes how metals are partitioned in various pools and how a rigorous extraction using HF is required to release metals present in the structure of primary or secondary minerals It is

3 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

expected that native metals (ie background concentrations) can be present in these structures The paper indicates that for four metals examined (cadmium copper zinc and nickel) the amount of metals that was bound in the mineral structure ranges from approximately 15-60 It would be expected that naturally occurring uranium would have similar characteristics

The National Institutes of Standards amp Technology (NIST) recognizes the differences between total digestion and method SW846-3050 and has issued Addendum to Certificates for the Certificate of Analysis (NIST 2003) for some Standard Reference Materials (SRMs) The Addendum to Certificates for SRM 2711 a Montana agricultural soil indicates that trace metals were measured at approximately 40 to 90 of their certified value when analyzed using SW846-3050 as the preparation method Even though uranium is not listed as a target analyte it is reasonable to expect that is would behave in a similar fashion

Sample Quanitation

The preparation method for alpha spec analysis includes addition of a tracer (U-232) to sample and is carried through the entire analytical scheme The amount of tracer is measured during analysis and the results for target compounds (the other uranium isotopes) are corrected for the amount of tracer recovery (yield) For example if amount of tracer measured is 50 of what was initially added to the sample the sample results are all corrected by dividing by 050 In terms of method quality assurancequality control (QAQC) the PRP QAPP indicates that tracer recovery in the range of 15-125 is considered acceptable and the analytical system is considered in control In other words up to 85 of the compound can be lost during sample preparation and analysis and the data is considered meeting all QAQC objectives This method of analysis is the equivalent to the technique of isotope dilution which is used for some non-radiological environmental analyses (eg EPA Method 1625 Semivolatile Organic Compounds by Isotope Dilution GCMS) Analyses using this technique although considered the most accurate are not widely employed due to the complexity and expense of using the isotope dilution technique

ICP-MS analysis measures the concentration of analyte with no correction for potential losses However QAQC procedures are in place to make sure the losses are not considered significant In addition to laboratory control samples matrix spike analysis measures the recovery for the target analyte that was over-spiked into the sample The PRP QAPP states that matrix spike recovery should be within the range of 75-125 Therefore up to 25 of the analyte could be lost and the method has still met the QAQC objectives If recovery is less than this then validation procedures acknowledge that the data may be biased low and sample results are flagged as estimated

Therefore because the alpha spec data is corrected for analyte loss during the analytical process while ICP-MS data is not some amount of positive bias is expected to be present in alpha spec data when compared to ICP-MS data

Considerations

In evaluating which analysis is appropriate for samples from the project neither method is inherently better than the other The following factors should be considered

1 What is being measured ICP-MS measures mass directly and data can be mathematically converted to activity Alpha spec measures activity directly and can be mathematically converted to mass

4 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

2 What are the objectives of the data in terms of measuring total versus environmentally available The ICP-MS method specified in the PRP QAPP measures uranium concentrations in the same manner that EPA has measured other metals concentrations Should uranium mass be measured in a typical way that will be used for measuring other metals of concern at the site (eg copper beryllium arsenic etc)

3 If ICP-MS is more cost effective that alpha spec should the option of using preparation method SW846-3052 be considered This should yield total uranium data similar to alpha spec analysis while still using the efficiencies of ICP-MS analysis

4 What units will the cleanup level for the site be in If its in mass units then ICPshyMS analysis may be more appropriate If its in activity units then alpha spec analysis may be warranted

References

Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts Technical Memorandum from Heath Downey to Bruce Thompson de maxim is inc September 26 2005 (MACTEC 2005)

Test Methods for Evaluating Solid Waste PhysicalChemical Methods SW-846 Third Edition with updates EPA 1996

Ambient Levels of Metals in New Jersey Soils Environmental Assessment and Risk Analysis Element Research Project Summary Division of Science Research and Technology State of New Jersey Department of Environmental Protection May2003

Ground Water Issue Behavior of Metals in Soil EPA Superfund Technology Support Center for Ground Water EPA540S-92018 October 1993

Certificate of Analysis Standard Reference Material 2711 Montana Soil National Institute of Standards amp Technology July 18 2003

bull bull

bull bull bullbull bull

10000

MampE SoilSediment Results Total Uranium mgKg

bull

bull - ~-~-----~~---~~-~--~~1000

en aE

I I bull Total a u

I -

bull bull 100 1shy -middot- -middot bullbull bull

Ybull bull

10 100 1000 10000

Alpha Spec

Attachm

ent 1

10 -

100000

TOTAL URANIUM BY ICP-MS mgKg

bull

bull 10000

I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 22: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

2 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

These two reasons are discussed in further detail as follows

Sample Preparation

Radionuclide analysis traditionally uses a sample preparation method that includes hydrofluoric acid (HF) to break down the sample so that all of the uranium present in the sample is put into solution On the other hand standard EPA methods for metals analysis including ICP-MS do not measure the total metals present in a sample For example method SW846-3050 (EPA 1996) (as specified by the PRP in their approved QAPP for metals analysis by ICP-MS) states

This method is not a total digestion technique for most samples It is a very strong acid digestion that will dissolve almost all elements that could become environmentally available By design elements bound in silicate structures are not normally dissolved by this procedure as they are not usually mobile in the environment If absolute total digestion is required use Method 3052

As noted by the method a SW846 method is available to do a total digestion (equivalent to the alpha spec method) as method 3052 However method 3052 (EPA 1996) states (emphasis added)

Note This technique is not appropriate for regulatory applications that require the use of leachate preparations (ie Method 3050 Method 3051 Method 1311 Method 1312 Method 1310 Method 1320 Method 1330 Method 3031 Method 3040) This method is appropriate for those applications requiring a total decomposition for research purposes (ie~ geological studiesr mass balancesr analysis of Standard Reference Materials) or in response to a regulation that requires total sample decomposition

Even though the ICP-MS preparation method does not yield total results it still may conservatively measure the amount of metals that are environmentally available In the Research Project Summary Ambient Levels of Metals in New Jersey Soils prepared by the New Jersey Department of Environmental Protection (NJDEP 2003) where samples were prepared using SW846-3050 procedures it states

Metal concentrations reported in this summary were acid-extractable metals not necessarily total metal concentrations The extraction method (USEPA Method 3050) is a vigorous extraction method designed to remove all metals that could become environmentally available In practice the extraction method likely overestimates metals that could become available since the sample is refluxed with both concentrated nitric acid and hydrogen peroxide However by design the method will not extract chemicals incorporated in silica minerals as they are usually not mobile in the environment Thus concentrations reported in these studies may be lower than those indicated from analyses using methods designed to measure total metal concentrations such as x-ray fluorescence methods

Although references to uranium were not located it is expected that uranium will have some of the same type of mobility characteristics as other metals in the soil matrix The paper Ground Water Issue Behavior of Metals in Soil published by EPAs Superfund Technology Support Center for Ground Water (EPA 1992) describes how metals are distributed within soil The paper describes how metals are partitioned in various pools and how a rigorous extraction using HF is required to release metals present in the structure of primary or secondary minerals It is

3 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

expected that native metals (ie background concentrations) can be present in these structures The paper indicates that for four metals examined (cadmium copper zinc and nickel) the amount of metals that was bound in the mineral structure ranges from approximately 15-60 It would be expected that naturally occurring uranium would have similar characteristics

The National Institutes of Standards amp Technology (NIST) recognizes the differences between total digestion and method SW846-3050 and has issued Addendum to Certificates for the Certificate of Analysis (NIST 2003) for some Standard Reference Materials (SRMs) The Addendum to Certificates for SRM 2711 a Montana agricultural soil indicates that trace metals were measured at approximately 40 to 90 of their certified value when analyzed using SW846-3050 as the preparation method Even though uranium is not listed as a target analyte it is reasonable to expect that is would behave in a similar fashion

Sample Quanitation

The preparation method for alpha spec analysis includes addition of a tracer (U-232) to sample and is carried through the entire analytical scheme The amount of tracer is measured during analysis and the results for target compounds (the other uranium isotopes) are corrected for the amount of tracer recovery (yield) For example if amount of tracer measured is 50 of what was initially added to the sample the sample results are all corrected by dividing by 050 In terms of method quality assurancequality control (QAQC) the PRP QAPP indicates that tracer recovery in the range of 15-125 is considered acceptable and the analytical system is considered in control In other words up to 85 of the compound can be lost during sample preparation and analysis and the data is considered meeting all QAQC objectives This method of analysis is the equivalent to the technique of isotope dilution which is used for some non-radiological environmental analyses (eg EPA Method 1625 Semivolatile Organic Compounds by Isotope Dilution GCMS) Analyses using this technique although considered the most accurate are not widely employed due to the complexity and expense of using the isotope dilution technique

ICP-MS analysis measures the concentration of analyte with no correction for potential losses However QAQC procedures are in place to make sure the losses are not considered significant In addition to laboratory control samples matrix spike analysis measures the recovery for the target analyte that was over-spiked into the sample The PRP QAPP states that matrix spike recovery should be within the range of 75-125 Therefore up to 25 of the analyte could be lost and the method has still met the QAQC objectives If recovery is less than this then validation procedures acknowledge that the data may be biased low and sample results are flagged as estimated

Therefore because the alpha spec data is corrected for analyte loss during the analytical process while ICP-MS data is not some amount of positive bias is expected to be present in alpha spec data when compared to ICP-MS data

Considerations

In evaluating which analysis is appropriate for samples from the project neither method is inherently better than the other The following factors should be considered

1 What is being measured ICP-MS measures mass directly and data can be mathematically converted to activity Alpha spec measures activity directly and can be mathematically converted to mass

4 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

2 What are the objectives of the data in terms of measuring total versus environmentally available The ICP-MS method specified in the PRP QAPP measures uranium concentrations in the same manner that EPA has measured other metals concentrations Should uranium mass be measured in a typical way that will be used for measuring other metals of concern at the site (eg copper beryllium arsenic etc)

3 If ICP-MS is more cost effective that alpha spec should the option of using preparation method SW846-3052 be considered This should yield total uranium data similar to alpha spec analysis while still using the efficiencies of ICP-MS analysis

4 What units will the cleanup level for the site be in If its in mass units then ICPshyMS analysis may be more appropriate If its in activity units then alpha spec analysis may be warranted

References

Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts Technical Memorandum from Heath Downey to Bruce Thompson de maxim is inc September 26 2005 (MACTEC 2005)

Test Methods for Evaluating Solid Waste PhysicalChemical Methods SW-846 Third Edition with updates EPA 1996

Ambient Levels of Metals in New Jersey Soils Environmental Assessment and Risk Analysis Element Research Project Summary Division of Science Research and Technology State of New Jersey Department of Environmental Protection May2003

Ground Water Issue Behavior of Metals in Soil EPA Superfund Technology Support Center for Ground Water EPA540S-92018 October 1993

Certificate of Analysis Standard Reference Material 2711 Montana Soil National Institute of Standards amp Technology July 18 2003

bull bull

bull bull bullbull bull

10000

MampE SoilSediment Results Total Uranium mgKg

bull

bull - ~-~-----~~---~~-~--~~1000

en aE

I I bull Total a u

I -

bull bull 100 1shy -middot- -middot bullbull bull

Ybull bull

10 100 1000 10000

Alpha Spec

Attachm

ent 1

10 -

100000

TOTAL URANIUM BY ICP-MS mgKg

bull

bull 10000

I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 23: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

3 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

expected that native metals (ie background concentrations) can be present in these structures The paper indicates that for four metals examined (cadmium copper zinc and nickel) the amount of metals that was bound in the mineral structure ranges from approximately 15-60 It would be expected that naturally occurring uranium would have similar characteristics

The National Institutes of Standards amp Technology (NIST) recognizes the differences between total digestion and method SW846-3050 and has issued Addendum to Certificates for the Certificate of Analysis (NIST 2003) for some Standard Reference Materials (SRMs) The Addendum to Certificates for SRM 2711 a Montana agricultural soil indicates that trace metals were measured at approximately 40 to 90 of their certified value when analyzed using SW846-3050 as the preparation method Even though uranium is not listed as a target analyte it is reasonable to expect that is would behave in a similar fashion

Sample Quanitation

The preparation method for alpha spec analysis includes addition of a tracer (U-232) to sample and is carried through the entire analytical scheme The amount of tracer is measured during analysis and the results for target compounds (the other uranium isotopes) are corrected for the amount of tracer recovery (yield) For example if amount of tracer measured is 50 of what was initially added to the sample the sample results are all corrected by dividing by 050 In terms of method quality assurancequality control (QAQC) the PRP QAPP indicates that tracer recovery in the range of 15-125 is considered acceptable and the analytical system is considered in control In other words up to 85 of the compound can be lost during sample preparation and analysis and the data is considered meeting all QAQC objectives This method of analysis is the equivalent to the technique of isotope dilution which is used for some non-radiological environmental analyses (eg EPA Method 1625 Semivolatile Organic Compounds by Isotope Dilution GCMS) Analyses using this technique although considered the most accurate are not widely employed due to the complexity and expense of using the isotope dilution technique

ICP-MS analysis measures the concentration of analyte with no correction for potential losses However QAQC procedures are in place to make sure the losses are not considered significant In addition to laboratory control samples matrix spike analysis measures the recovery for the target analyte that was over-spiked into the sample The PRP QAPP states that matrix spike recovery should be within the range of 75-125 Therefore up to 25 of the analyte could be lost and the method has still met the QAQC objectives If recovery is less than this then validation procedures acknowledge that the data may be biased low and sample results are flagged as estimated

Therefore because the alpha spec data is corrected for analyte loss during the analytical process while ICP-MS data is not some amount of positive bias is expected to be present in alpha spec data when compared to ICP-MS data

Considerations

In evaluating which analysis is appropriate for samples from the project neither method is inherently better than the other The following factors should be considered

1 What is being measured ICP-MS measures mass directly and data can be mathematically converted to activity Alpha spec measures activity directly and can be mathematically converted to mass

4 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

2 What are the objectives of the data in terms of measuring total versus environmentally available The ICP-MS method specified in the PRP QAPP measures uranium concentrations in the same manner that EPA has measured other metals concentrations Should uranium mass be measured in a typical way that will be used for measuring other metals of concern at the site (eg copper beryllium arsenic etc)

3 If ICP-MS is more cost effective that alpha spec should the option of using preparation method SW846-3052 be considered This should yield total uranium data similar to alpha spec analysis while still using the efficiencies of ICP-MS analysis

4 What units will the cleanup level for the site be in If its in mass units then ICPshyMS analysis may be more appropriate If its in activity units then alpha spec analysis may be warranted

References

Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts Technical Memorandum from Heath Downey to Bruce Thompson de maxim is inc September 26 2005 (MACTEC 2005)

Test Methods for Evaluating Solid Waste PhysicalChemical Methods SW-846 Third Edition with updates EPA 1996

Ambient Levels of Metals in New Jersey Soils Environmental Assessment and Risk Analysis Element Research Project Summary Division of Science Research and Technology State of New Jersey Department of Environmental Protection May2003

Ground Water Issue Behavior of Metals in Soil EPA Superfund Technology Support Center for Ground Water EPA540S-92018 October 1993

Certificate of Analysis Standard Reference Material 2711 Montana Soil National Institute of Standards amp Technology July 18 2003

bull bull

bull bull bullbull bull

10000

MampE SoilSediment Results Total Uranium mgKg

bull

bull - ~-~-----~~---~~-~--~~1000

en aE

I I bull Total a u

I -

bull bull 100 1shy -middot- -middot bullbull bull

Ybull bull

10 100 1000 10000

Alpha Spec

Attachm

ent 1

10 -

100000

TOTAL URANIUM BY ICP-MS mgKg

bull

bull 10000

I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 24: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

4 Attachment 1Uranium Analysis comparison of alpha spectrometry and ICP-MS results

2 What are the objectives of the data in terms of measuring total versus environmentally available The ICP-MS method specified in the PRP QAPP measures uranium concentrations in the same manner that EPA has measured other metals concentrations Should uranium mass be measured in a typical way that will be used for measuring other metals of concern at the site (eg copper beryllium arsenic etc)

3 If ICP-MS is more cost effective that alpha spec should the option of using preparation method SW846-3052 be considered This should yield total uranium data similar to alpha spec analysis while still using the efficiencies of ICP-MS analysis

4 What units will the cleanup level for the site be in If its in mass units then ICPshyMS analysis may be more appropriate If its in activity units then alpha spec analysis may be warranted

References

Radiological Review of Fall 2004 Rl Analytical Data Nuclear Metals Superfund Site Concord Massachusetts Technical Memorandum from Heath Downey to Bruce Thompson de maxim is inc September 26 2005 (MACTEC 2005)

Test Methods for Evaluating Solid Waste PhysicalChemical Methods SW-846 Third Edition with updates EPA 1996

Ambient Levels of Metals in New Jersey Soils Environmental Assessment and Risk Analysis Element Research Project Summary Division of Science Research and Technology State of New Jersey Department of Environmental Protection May2003

Ground Water Issue Behavior of Metals in Soil EPA Superfund Technology Support Center for Ground Water EPA540S-92018 October 1993

Certificate of Analysis Standard Reference Material 2711 Montana Soil National Institute of Standards amp Technology July 18 2003

bull bull

bull bull bullbull bull

10000

MampE SoilSediment Results Total Uranium mgKg

bull

bull - ~-~-----~~---~~-~--~~1000

en aE

I I bull Total a u

I -

bull bull 100 1shy -middot- -middot bullbull bull

Ybull bull

10 100 1000 10000

Alpha Spec

Attachm

ent 1

10 -

100000

TOTAL URANIUM BY ICP-MS mgKg

bull

bull 10000

I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 25: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

bull bull

bull bull bullbull bull

10000

MampE SoilSediment Results Total Uranium mgKg

bull

bull - ~-~-----~~---~~-~--~~1000

en aE

I I bull Total a u

I -

bull bull 100 1shy -middot- -middot bullbull bull

Ybull bull

10 100 1000 10000

Alpha Spec

Attachm

ent 1

10 -

100000

TOTAL URANIUM BY ICP-MS mgKg

bull

bull 10000

I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 26: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

100000

TOTAL URANIUM BY ICP-MS mgKg

bull

bull 10000

I

bullra +- ra I bull bull

- 0 1000a I0 bull

a

laquobull bull bull

01 10 100 1000

MampE Data

Attachm

ent 1

100 bull pound

010

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 27: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS

(GL-MA-E-009 REVISION 10)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or otherwise used for the benefit of others except by express written permission of GEL

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 28: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Acid Digestion of Sediments Sludges and Soils Attachment 2

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 2 of 8

TABLE OF CONTENTS

10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS SLUDGES AND SOILS 3

20 PURPOSE 3

30 DISCUSSION 3

40 DEFINITIONS 3

50 PROCEDURES 4

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS 5

70 RECORDS MANAGEMENT 6

81 LABORATORY WASTE 7

90 REFERENCES 7

APPENDIX 1 SAMPLE PREP LOGBOOK 8

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 29: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 3 of 8 10 STANDARD OPERATING PROCEDURE FOR THE ACID DIGESTION OF SEDIMENTS

SLUDGES AND SOILS 20 PURPOSE

To describe the manner in which sediments sludges and soils for Inductively Coupled Plasma (ICP) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS) analysis are digested by EPA SW-846 Method 3050B Samples digested by this procedure are applicable for analysis by SW-846 methods

30 DISCUSSION A representative portion of sample is digested with nitric acid and hydrogen peroxide This digestate is then refluxed with hydrochloric acid for ICP analysis only This procedure deviates from Method 3050B in that sample volumes are half the method recommendations

40 DEFINITIONS 41 Blank - Type I water that has been taken through the digestion process The

blank is used to determine the amount of background contamination 42 Laboratory Control Sample (LCS) - A certified reference material that has been

taken through the digestion process The LCS is used to determine digestion accuracy and to determine if the digestion process is in control

43 Laboratory Control Sample Duplicate (LCS DUP) - A duplicate of the LCS The LCS DUP is used to determine reproducibility and to indicate precision

44 Matrix Spike (MS) - A sample that has added to it a known amount of solution containing known concentrations of analytes The MS is used to determine the presence or absence of interferences and matrix effects in the digested sample

45 Matrix Spike Duplicate (MSD) - A duplicate of the MS The MSD indicates reproducibility

46 Sample Duplicate (DUP) - A duplicate of a sample The DUP indicates reproducibility

47 Type II water - Water that conforms to the following performance specifications Electrical resistivity min 10 Total organic carbon max microL 50 Sodium max microgL 5 Chlorides max microgL 5 Total silica max microgL 3

Type II water is dispensed throughout the laboratory through a centralized distribution system

48 Type I water - Water that conforms to the following performance specifications Electrical resistivity min 1667 Total organic carbon max microgL 100 Sodium max microgL 1 Chlorides max microgL 1 Total silica max microgL 3

Type I water is dispensed within the metals prep lab by the ldquoMilliQrdquo water system 49 11 HNO3 - Concentrated reagent grade nitric acid that has been mixed with Type

I water in a ratio of one to one 410 HNO3 - Concentrated reagent grade 709 nitric acid

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 30: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 4 of 8

411 HCl - Concentrated reagent grade 37 hydrochloric acid 412 H2O2 - Concentrated 30 hydrogen peroxide

50 PROCEDURES A batch consists of samples of the same matrix and quality control (QC) samples that are digested together Each of the quality control samples listed in section 41 through 46 must be included in each batch at the frequency listed or as per client request The blank LCS andor LCS DUP are digested at a frequency of one in 20 or per batch whichever is more frequent The MS MSD andor DUP are digested at a frequency of one in 20 or per batch whichever is more frequent or per specified clientprogram requirements 51 Glassware preparation

511 Glassware that has been cleaned according to Glassware Preparation (GLshyLB-E-003) is soaked in a water and acid mixture for at least 30 minutes

512 After soaking the glassware is rinsed with copious quantities of Type II or Type I water and then inverted over clean absorbent paper or onto a rack for drying

52 Label the Teflon beakers or centrifuge tube with the sample numbers in the batch If centrifuge tube is to be used it must be calibrated before usage See SOP GLshyLB-E-026 for centrifuge testing procedure

53 Mix the sample to achieve homogeneity Weigh approximately 05 to 510 grams of sample for ICPMS and 1 to 110 grams if sample is for ICPAES Transfer the weighed sample to the appropriately labeled Teflon beaker or centrifuge tube 531 Sample aliquots should not be taken from the top of an unmixed sample

because large particles tend to rise in solid matrixes and heavy materials tend to sink in liquid matrixes

532 Powdered samples may be homogenized by gently rocking the sample side to side Then a representative aliquot may be taken from the center of the powder

533 Other matrixes must be stirred turned or mixed before sampling 54 Quality control samples are prepared prior to digestion

541 The beaker or tube to be used for the blank MS MSD andor DUP LCS andor LCS DUP is labeled

542 Weigh approximately 05 to 100 gram of sample and transfer to the MS MSD andor DUP beaker or tube 5421 The MS MSD LCS andor LCS Dup are spiked with known

amounts of spiking solution 543 Select a LCS Mix the LCS to achieve homogeneity Weigh approximately

05 to 100 gram of the sample and transfer to the LCS andor LCS DUP Teflon beaker or centrifuge tube

544 The blank beaker or tube is labeled and no water spike or sample is added to it

55 Add 5 mL of 11 HNO3 to the samples and quality control samples 56 Gently swirl the sample acid mixture 57 Cover the sample with a watch glass and heat the sample on a hot plateblock to

near boiling Reflux the sample for 10 to 15 minutes GENERAL ENGINEERING LABORATORIES LLC

PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 31: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 5 of 8

58 Remove the sample from the hot plate or block and allow the sample to cool 59 Add 25 mL of concentrated HNO3 replace the watch glass and reflux for 30 mins 510 Using a ribbed watchglass or vapor recovery system allow the solution to

evaporate to approximately 05 mL without boiling or heat for 2 hours 5101 Remove the sample from the hot plate or block and allow the sample to

cool 511 Add 15 mL of H2O2 and 10 mL of Type I water Return the sample to the hot

plate or block and allow the peroxide reaction to occur Continue to add H2O2 to the sample until the effervescence subsides Do not add more than 5 mL H2O2 5111 Cover the sample with a ribbed watchglass heating the acid-peroxide

digestate until the volume is reduced to approximately 25 mL or heat at 95deg C plusmn 5deg C without boiling for 2 hours

512 If the sample is being prepared for ICP 5121 Remove the sample from the hot plate or block Allow sample to cool 5122 Add 5 mL of concentrated HCl and 25 mL of Type I water Return the

covered beaker or tube to the hot plate or block and reflux without boiling for 15 minutes

5123 Remove the sample from the hot plateblock 5124 Allow the sample to cool 5125 Dilute the sample to 50 mL with Type I water 5126 Cap and shake the sample 5127 Organize the samples in a storage container and label the container with the

batch number of the sample group 513 If the sample is being prepared for ICP-MS analysis

5131 Allow the sample to continue refluxing on the hot plate or block for at least 15 minutes or until the volume is reduced to 25 mL

5132 Do not allow the sample to go to dryness 5133 Remove the sample from the hot plate or block 5134 Allow the sample to cool 5135 Dilute the sample to 50 mL with Type I water 5136 Cap and shake the sample 5137 Filter each sample with a 20 microm pore size plunger type filter (PTF grade) or

allow to settle overnight 5138 Organize the samples in a storage container and label the container with the

batch number of the sample group 514 If the sample contains particulate material that could clog the nebulizer if

necessary you may filter or centrifuge the sample 55141 Be advised that filtration is a common cause of contamination If

a sample is filtered any QC associated with the sample must also be filtered Additionally if any sample in the batch is filtered the method blank and laboratory control sample must also be filtered

60 SAFETY HEALTH AND ENVIRONMENTAL HAZARDS WARNING

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 32: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 6 of 8

CONCENTRATED HCl AND HNO3 ARE EXTREMELY CORROSIVE AND CAN CAUSE SEVERE BURNS TO THE SKIN CONCENTRATED 30 H2O2 IS A VIOLENT OXIDIZER KEEP AWAY FROM OPEN FLAMES AND RINSE WITH WATER IF SKIN CONTACT OCCURS 61 Wear eye protection with side shields while performing procedures in the lab 62 Treat all chemicals and samples as a potential health hazard and reduce exposure

to these chemicals to the lowest level possible GEL maintains a current awareness file of OSHA regulations regarding the safe handling of the chemicals A reference file of Material Safety Data Sheets (MSDS) and individual client sample MSDSs are also maintained

63 Personal protective equipment 631 Disposable gloves are worn and changed frequently when working with

acids glassware or samples Dirty gloves pose a contamination hazard to the samples Gloves that have holes can be dangerous to the wearer by allowing acids and toxic metals to come in contact with skin

632 Hood doors are pulled down partially while digesting samples Acidified samples can splash and pop as they are being heated

633 To protect clothes and skin from exposure to corrosive material wear a lab jacket

64 Prior to handling radioactive samples analysts must have had radiation safety training and understand their full responsibilities in radioactive sample handling Some general guidelines follow 641 Wear plastic apron over lab coat when working with radioactive samples 642 Protect counter tops with counter paper or work from radioactive sample

handling trays 643 Prohibit admittance to immediate work area 644 Post signs indicating radioactive samples are in the area 645 Take swipes of the counter tops upon completion of work Deliver those

swipes to the designated swipe count box 646 Segregate radioactive wastes Radioactive waste containers are obtained

from the Waste Management 65 All samples chemicals extracts and extraction residues must be transferred

delivered and disposed of safely according to all related SOPs 651 Segregate solid wastes from liquid wastes in the satellite area containers 652 Segregate oil wastes from water-soluble wastes in the satellite area containers

66 In the event of an accident or medical emergency call for help immediately When time and safety permit an accident report form should be completed and turned in to the safety committee

67 Fire escape routes are posted in the lab and all personnel should be familiar with them In addition fire safety equipment such as fire extinguishers is located in the lab Training is available on the proper operation of this equipment

70 RECORDS MANAGEMENT 71 Upon completion of batch preparation digestion data shall be entered into the

LIMS Prep Logbook (see Appendix 1) following the guidelines in GL-LB-E-008

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 33: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 7 of 8

for ldquoBasic Requirements for the Use and Maintenance of Laboratory Notebooks Logbooks Forms and Other Record Keeping Devicesrdquo

72 Data to be entered into the logbook includes analyst name prep data and time initial volume or weight with units and final volume with units

73 Standards and reagents may also be entered into the logbook and fall under the guidelines of GL-LB-E-015 for ldquoControl of Laboratory Standardsrdquo and GL-LBshyE-007 for ldquoLaboratory Standards Documentationrdquo

74 Upon entry of prep data obtain a printout of the logbook The analyst listed on the logbook should sign and date the page near their printed initials The logbook page is kept with the samples with which it is associated

75 The entry of correct prep data is peer reviewed (correct dates times weights volumes SOPrevision spikes spike amounts and reagent information etc) Once data is reviewed the batch is statused to DONE in LIMS the logbook is signed and dated by the reviewer and the batch is ready for analysis

81 LABORATORY WASTE For the proper disposal of sample and reagent wastes from this procedure refer to the Laboratory Waste Management Plan GL-LB-G-001

90 REFERENCES 91 Test Method for Evaluating Solid Waste Laboratory Manual Physical Chemical

Methods Method 3050B Revision 2 December 1996 92 1992 Annual Book of ASTM Standards Standard D1193-91 ldquoStandard

Specification for Reagent Waterrdquo

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 34: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Prep LogBook

mI)JI J~( fypt S~Jitlol llaquolol Spilealmnl ipibU illl

IWh LCS J((((lllIS2 25 ml lbull~gtpDa-c (IJ]1JIt~1middot13ll LCS J(((OOlIS2 - 25 llll

Jib SOP~ lta bull ~uJBJl MS J((((GISS 25 bullbullll MS J((((QlISS - 25 ml

MSI) J~ISI 25 bullbullll MSI) JtumiSI - 25 loll

Type SJ~jlltvl l~lll111n(lto bull tolbod IIUWJ1 bull IIIIIIIV(IIIIIIIC IIIJgtIltlnr ((IIIIIIIIIIS MIIIIU

Mn J~ISJ 10022131 FuUIJsl FlaquoQC iflul ifhul Woller

SMJIII~ lloi~J 100 bull 22-3l7gt1coibullllll ifhul iflull W11s1Wuhr

LCS JOOOOJlIS2 1002Mll7 FuUIisl FlaquoQC ifhnl iflml Wulor

SMPIH 21lt116002 10022-117lcoiollll iflnll ifhnl l)osloWIIhr

SMPIH 21middot1~ 10022317 lcoiolll iflllll iflllll WsloWulor

SMPIH 21middot1~1 10022-3ll7 lcoi11m ifhnl iflml Wu~1oW11hr

Sl))ll J(((l2lISl 21gt116001 10022-lll71uu lisl Flaquo QC ifhul ifhnl wn r MSI) 100002lISI 21o116001 1tn223l71uulisl FoQC iflull iflllll Wuhr

MS 1~1SS 21middot11-SWI 10022lll71uu Usl FlaquoQC iflull iflnll Wnhr

Pbull~ct __ _

Attachment 2Acid Digestion of Sediments Sludges and Soils

SOP Effective 893 GL-MA-E-009-REV 10 Revision 10 Effective July 2003 Page 8 of 8

APPENDIX 1 SAMPLE PREP LOGBOOK

(for illustrative purposes only)

GENERAL ENGINEERING LABORATORIES LLC PO BOX 30712 Charleston SC 29417

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 35: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 2STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 1 of 8

VERIFY THE VALIDITY OF THIS SOP EACH DAY IN USE

STANDARD OPERATING PROCEDURE

FOR

DIGESTION FOR SOIL

(GL-RAD-A-015 REVISION 6)

PROPRIETARY INFORMATION

This document contains proprietary information that is the exclusive property of General Engineering Laboratories LLC (GEL) No contents of this document may be reproduced or used for the benefit of others except by written permission of GEL

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 36: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 2 of 8

TABLE OF CONTENTS

SECTION Page

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 3

20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY3

30 METHOD APPLICABILITY3

40 DEFINITIONS3

50 METHOD VARIATIONS 3

60 SAFETY PRECAUTIONS AND WARNINGS3

70 INTERFERENCES3

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION4

90 SAMPLE HANDLING AND PRESERVATION 4

100 SAMPLE PREPARATION 4

110 PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS 5

120 INSTRUMENT CALIBRATION AND PERFORMANCE5

130 ANALYSIS PROCEDURES AND INSTRUMENT OPERATION 5

140 EQUIPMENT AND INSTRUMENT MAINTENANCE5

150 DATA RECORDING CALCULATION AND REDUCTION METHODS 6

160 QUALITY CONTROL REQUIREMENTS 6

170 DATA REVIEW APPROVAL AND TRANSMITTAL 6

180 RECORDS MANAGEMENT 6

190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL 6

200 REFERENCES 6

APPENDIX 1 7

APPENDIX 2 8

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 37: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 3 of 8

10 STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL 20 METHOD OBJECTIVE PURPOSE CODE AND SUMMARY

21 This standard operating procedure provides the necessary instructions to conduct digestion on soil type matrices

22 This procedure is applicable to soils for any analysis requiring complete solution Sample aliquots of one gram or less can be accommodated by the equipment and amount of reagents specified Sequential analysis of more than one radionuclide may be accomplished by the addition of carriers and tracers for each applicable analysis prior to digestion

24 To minimize errors due to non-homogeneity of a sample it is always desirable to use the largest possible aliquot of sample The existence of ldquohot particlesrdquo may lead to questionable data when very small sample aliquots are used In cases where radionuclide concentrations are ldquohigher than normal environmental levelsrdquo and smaller aliquots are necessary it is preferable to dilute the digested 1-gram sample rather than select a smaller aliquot The analyst should consult the group leader for guidance in these situations

30 METHOD APPLICABILITY Method Detection Limit (MDL) See specific SOP for minimum detectable activity (MDA) values

40 DEFINITIONS 41 National Institute of Standards and Technology (NIST) For the purpose of this

method the national scientific body responsible for the standardization and acceptability of analyte solutions

42 Type II water Deionized (DI) water 50 METHOD VARIATIONS

Some variations may be necessary due to special matrices encountered in the lab These variations may be used with approval from a Group Leader or Senior Technical Specialist Variations to a method will be documented with the analytical raw data

60 SAFETY PRECAUTIONS AND WARNINGS 61 Personnel performing this analytical procedure are trained to the safe laboratory

practices outlined in the Safety Health and Chemical Hygiene Plan GL-LB-N-001 62 Personnel handling radioactive materials are trained in and follow the procedures

outlined in GL-RAD-S-004 for Radioactive Material Handling 63 Personnel handling biological materials are trained in and follow the procedures

outlined in GL-RAD-S-010 for Handling Biological Materials 64 If there is any question regarding the safety of any laboratory practice stop

immediately and consult qualified senior personnel such as a Group or Team Leader

70 INTERFERENCES Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 38: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 4 of 8

80 APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION 81 Teflon beakers 82 Reagents Chemicals and Standards

821 Nitric acid reagent grade (16 M) 822 Hydrofluoric acid 48 823 Hydrochloric acid reagent grade (12 M) 824 Type II water 825 Boric acid 5 Dissolve 50 grams of H3 BO3 per liter of water 826 Ammonium Hydroxide (NH4OH)

90 SAMPLE HANDLING AND PRESERVATION Soil and Sand require no preservation and may be shipped in any suitable container

100 SAMPLE PREPARATION 101 For instructions on drying grinding homogenizing and blending soil samples

refer to ldquoSoil Sample Preparation for the Determination of Radionuclidesrdquo (GLshyRAD-A-021)

102 Alpha Spec soil digest procedure 1021 Weigh an appropriate aliquot (05 ndash 10 grams) of ashed soil sample

into a Teflon beaker Correct the sample mass for losses due to ashing and record the corrected weight on the Que sheet

1022 Add appropriate tracers and record tracer IDs and tracer volumes on Que sheet

1023 Add 10 mL of 48 HF to each Teflon beaker Cover beakers with a Teflon cover and place on hotplate on medium heat for 30 minutes Remove lids and evaporate solution to dryness

1024 Add 10 mL of concentrated nitric acid and 10 mL of 48 HF to each Teflonreg beaker Cover beakers with a Teflonreg cover and place on a medium heat for 30 minutes Remove lids and evaporate the to solution to dryness

1025 Add 10 mL of concentrated HC1 and 5 mL of 48 HF to each sample cover with Teflon lid and heat for 30 minutes on the hotplate Remove the lid and evaporate to dryness on a hotplate

1026 Add 10 mL concentrated HC1 and 10 mL concentrated nitric acid to each Teflon beaker and evaporate to dryness on a hotplate

1027 Add 10 mL concentrated HC1 and 1 mL of saturated boric acid to each sample and evaporate to dryness on a hotplate set on medium heat

10271 Add 5 mL concentrated HNO3 dry (ThU and Ra samples only)

1028 If further clean up of sample is necessary continue with step 1029 Otherwise proceed to step 10211

1029 Dissolve the sample residue in 10 ndash 15 mL of concentrated hydochloric acid and transfer to a disposable centrifuge tube using DI

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 39: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 5 of 8

110

120

130

140

water as a rinse Add concentrated ammonium hydroxide until iron hydroxide precipitates then add 2 mL excess NH4OH Centrifuge sample and discard supernate (Do not do this if ThU or Ra only AmCm Pu U batches)

10210 Wash precipitate with 20 mL of DI water that has been adjusted to pH 10 with amm onium hydroxide Centrifuge sample and discard supernate

10211 For thorium analysis dissolve iron precipitate in 15 m L of 8 M nitric acid cetnrifuge sample and proceed to column work

10212 For any combination of analyses requiring isotopic U Pu Np or AmCm diss olve the iron precipitate in the required reagent based on the isotope

10213 For Radium-226 analysis pro ceed to GL-RAD-A-046 Step 104 103 Gross AlphaBeta soil digest procedure

1031 Weigh out an appropriate aliquot (nor mally 01g) into a teflon beaker Record this weight on the que sheet

1032 Add spike solu tions to the applicable samples and record volumes on the que sheet

1033 Add 10 mL of concentrated nitric acid to each sample 1034 Place samples on medium heat (sim300 degF) and co ver each sample with

a teflon lid Reflux all samples for 30 minutes 1035 Remove teflon lids and add 5 mL concentrated hydrochloric acid and

10 mL hydrof luoric acid to each sample Cover samples and reflux for 120 minutes

1036 Remove teflon lids and allow samples to evaporate to dryness 1037 Add 5 mL of concen trated nitric acid and evaporate to dryness 1038 Repeat Step 1037 1039 Add 5 mL of concentrated nitric acid to the dry samples Place the

samples back on the hot plate long enough so that the dried sample dissolves into the acid

10310 Samples are now ready for plancheting PREPARATIO N OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS Refer to ldquoPreparation of Radioactive Standardsrdquo (GL-RA D-M-001) INSTRUMENT C ALIBRATION AND PERFORMANCE Not Applicable ANALYSIS PRO CEDURES AND INSTRUMENT OPERATION Not Applicable EQUIPMENT A ND INSTRUMENT MAINTENANCE Not Applicable

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 40: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 6 of 8

150 DATA RECORDING CALCULATION AND REDUCTION METHODS Not Applicable

160 QUALITY CONTROL REQUIREMENTS 161 Analyst and Method Verification

Refer to GL-RAD-D-002 ldquoAnalytical Methods Validation for Radiochemistryrdquo for instructions concerning the validation of analysts and analytical methods

162 Method Specific Quality Control Requirements See specific SOP for specific method quality control requirements

170 DATA REVIEW APPROVAL AND TRANSMITTAL Not Applicable

180 RECORDS MANAGEMENT 181 Each analysis that is performed on an instrument is documented in the run log

according to ldquoRun Logsrdquo GL-LB-E-009 182 All raw data printouts calculation spreadsheets and batch checklists are filed with

the sample data for archival and review 190 LABORATORY WASTE HANDLING AND WASTE DISPOSAL

Radioactive samples and material are be handled and disposed in accordance with the Laboratory Waste Management Plan GL-LB-G-001

200 REFERENCES None

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 41: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 7 of 8

APPENDIX 1

ALPHA SPECTROSCOPY SOIL DIGESTION

____ Weigh appropriate amount of ashed soil to meet MDA

____ Add appropriate tracers and spikes to samples

____ 10 mL[HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HCl] and 5 mL [HF]

____ Reflux for 30 minutes

____ Evaporate to dryness

____ 10 mL [HNO3] and 10 mL [HCl]

____ Evaporate to dryness

____ 10 mL [HCl] and 1 mL Boric Acid

___ _ Evaporate to dryness

___ _ For Thorium or Radium samples add 5 mL [HNO ] and evaporate to dryness 3

____ Proceed to the appropriate procedure for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 42: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____ ____

STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL Attachment 2

SOP Effective Date 120892 GL-RAD-A-015 Rev 6 Revision 6 Effective September 2003 Page 8 of 8

APPENDIX 2

GROSS ALPHABETA SOIL DIGESTION

Weigh appropriate amount of dry soil to meet MDA

Add appropriate spikes to samples

10 mL [HNO3]

Reflux for 30 minutes

5 mL [HCl] and 10 mL [HF]

Reflux for 120 minutes

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

5 mL [HNO3]

Evaporate to dryness

Dissolve sample with 5 mL [HNO3]

Proceed to GL-RAD-A-001B for completion of analysis

General Engineering Laboratories LLC PO Box 30712 Charleston SC 29417

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 43: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 3

DEPARTMENT OF THE ARMY US ARMY TEST MEASUREMENT AND DIAGNOSTIC EQUIPMENT ACTIVITY

PROJECT DEVELOPMENT AND RADIATION RESEARCH OFFICE ATTN AMSAMndashTMD-SB

5905 PUTNAM ROAD SUITE 1 FORT BELVOIR VIRGINIA 22060-5448

REPLY TO ATTENTION OF

AMSAM-TMD-SB January 19 2000

MEMORANDUM THRU Commander US Army Materiel Command ATTN AMCSF (Mr Pittenger) 5001 Eisenhower Avenue Alexandria VA 22333-0001

THRU Commander US Army Materiel Command ATTN AMCSG-R (LTC Melanson) 5001 Eisenhower Avenue Alexandria VA 22333-0001

FOR Commander US Tank and Automotive Command ATTN AMSTA-CM-PS (Ms Lapajenko-Maguire) Warren MI 48397

SUBJECT Analysis of Transuranics and Other Contaminants in Depleted Uranium Armor

1 Reference memorandum Headquarters US Army Materiel Command Safety Office AMCSF-P dated 7 October 1999 subject Designation of Project Officer

2 IAW the above reference an analysis was made of the levels of transuranic elements (TRUS) and fission products (FP) in depleted uranium armor

3 The report is included as enclosure 1 and was completed with the assistance of the US Army Heavy Metals Office and the US Army Center for Health Promotion and Preventive Medicine

4 The levels of TRUs and FPs are in the pCig range and result in less than a one percent increase in dose as represented by the Annual Limits on Intake Hence the presence of these trace amounts of contaminants in DU armor is safe

5 POC for this matter is the undersigned at DSN 654-1979 Comm (703)-704-1979 or by e-mail rbhatbelvoirarmymil

RAMACHANDRA K BHAT PhD CHP Chief Project Development and Radiation Research Office USATA AMCOM

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 44: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 3

Enclosure 1 Tank-Automotive and Armaments Command (TACOM) and Army Materiel Command (AMC) Review of Transuranics (TRU) in Depleted Uranium (DU)

Armor

Project Officer Ramachandra K Bhat PhD CHP

January 19 2000

BACKGROUND

The United States Department of the Army (DA) Nuclear Regulatory Commission (NRC) license for Depleted Uranium (DU) armor is managed by the Tank-automotive and Armaments Command (TACOM)

In August 1999 the Department of Energy (DOE) informed the Army that DU armor carried trace amounts of transuranics (TRU) and Technetium-99 (Tc-99) The NRC subsequently requested the Army submit an amendment to its DU Armor license (NRC Materials License No SUB-1536 Docket No 040shy08994) to reflect the quantities of TRU contained in the Armyrsquos DU Armor based on more extensive samplinglab analysis To provide the NRC with such analysis the Army Materiel Command (AMC) developed a plan designed to quantitatively assess TRU content in DU Armor

QUALITY ASSURANCE (QA)

According to ANSIHPS N131-1999 quality assurance includes ldquoplanned and systematic actions necessary to provide confidence that a system or component will perform satisfactorily in service and that the results are both correct and traceablerdquo

Dr Bhat was tasked by AMC headquarters as Project Officer of the Analysis of Transuranics in Depleted Uranium Project The projectrsquos goals are to assess levels of TRU in DU for the Armyrsquos DU Armor license and characterize the risk in terms of relative increase in Annual Limits on Intake (ALI) In order to accomplish this objective Dr Bhat consulted with the NRC license holder and Army agencies including the US Army Center for Health Promotion and Preventive Medicine (USA CHPPM) Collectively the agencies designed a quality assurance program establishing guidelines to be followed by the designated laboratories during the analysis of TRU in DU Armor Highlights of the established criteria are listed below Each selected laboratory should have

- An established performance record in DOE Quality Assurance Program (QAP)Mixed Analyte Performance Evaluation Program (MAPEP) -The capability to analyze spiked samples to check for both chemical and radiological accuracy prior to sample analysis -The capability to analyze one spiked TRU in uranium sample prior to sample analysis -The capability to obtain a Minimum Detectable Concentration (MDC) of 1 pCi of TRUg of DU -Laboratory procedures which are well established and published in the literature

2

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 45: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 3

Dr Ramachandra Bhat traveled to the Idaho National Engineering and Environmental Laboratory (INEEL) to visit the Idaho Nuclear Technology and Engineering Center (INTEC) and the Radiological and Environmental Science Laboratory (RESL) Dr Bhat reviewed INTECrsquos established quality assurance program and subsequently determined that their assistance would be beneficial to the DU Armor project and designated INTEC as the primary laboratory A secondary laboratory RESL (RESL evaluates the performance of DOE laboratories by administering a QC program called MAPEP) was designated as an additional laboratory to verify the analytical performance of INTEC the primary laboratory QC measures established by RESL include the production of TRU spiked standards in a uranium matrix for the performance evaluation of INTEC

As a QC measure Dr Bhat requested that 10 of the samples analyzed by INTEC be analyzed by RESL Dr Bhat established the Video Teleconference (VTC) format as the forum for a collaborative decision-making process involving participants from DOE INTEC RESL DA and Air Force officials (who were observers) VTCs took place in November and December 1999 and January 2000 In each VTC session QA was granted the highest priority to obtain the credibility of the TRU in DU Armor values reported by INTEC

SAMPLE PREPARATION AND LOGISTICS

DOE shipped DU billets to a contractor Specific Manufacturing Capability (SMC) located in Idaho Falls Idaho SMC produced the DU armor from DU billets

DU armor is shipped from SMC to Lima Tank Plant for insertion into tanks SMC ships the scraps from armor production to another contractor Starmet Inc in Boston MA This is then melted and recast into billets and is then sent back to SMC During recycling of nuclear fuel TRUs and long-lived fission products entered the DU stream

The Army decided to analyze both quantitatively and qualitatively random samples from three different generations or populations of finished billets Population 1 is comprised of billets from the original shipment of DU Armor Scraps from the production process are melted and recast into billets Population 2 contains billets recast from Population 1 Population 3 contains billets recast from the production of Population 2 This process of recasting scrap is the reason no additional DU Armor has been added to the process since the first shipment

At the request of TACOM DOE prepared three sets of sixty billet samples Samples from Population 1 billets were taken at SMC Two samples were taken from each billet selected one inch removed from the edges of the long face of each billet Two samples one from each end of the billet were taken to assess the homogeneity of the billet The samples were obtained by drilling at an approximately 1rdquo depth and collecting 40g of DU turnings or shavings per sample Because DU is highly pyrophoric the drilling had to be done with the block submerged in a coolant comprised of water and Trimor A fresh drill bit was used for each end of each block to eliminate the possibility of cross-contamination Starmet Inc archived one-inch cubes taken from the top crop of Population 2 and 3 billets The selected cubes were sent to SMC for sampling The sampling of these cubes was performed as described above for Population 1 billet samples When the three sets of sixty were completed SMC shipped one set of 60 samples to INTEC the second set was designated for AMC and the third set was put into storage for future research

3

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 46: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 3

SAMPLE ANALYSIS STRATEGY

The sixty samples designated for INTEC are the primary focus of this study The samples were comprised of 20 Population 1 (the first population was the initial 12 million pounds of DU Armor billets cast at Fernald) billet samples 30 samples from Population 2 (the second population is the 2992 first recycle billets) billets and 10 samples from Population 3 (the third population is the second recycle billets which wereare produced from recycling first recycle scrap) billets The Armyrsquos sampling strategy was designed to simultaneously create a statistically representative sampling of the DU Armor production lot from 1986 to present and to be cost effective In order to accomplish this objective a professional statistician was consulted to select 60 total samples from 3 populations1 After selection was complete the samples were distributed among two separate laboratories INTEC the primary laboratory employed received and analyzed all 60 samples RESL a DOE laboratory received 8 out of the 60 total samples allocated for AMC Quality Assurance (QA) and Quality Control (QC) information were collected from INTEC along with the results of sample analysis2 The QAQC data collected allow for a realistic interpretation of the sample analysis results The accuracy of each laboratoryrsquos analysis may be determined from its long-term performance in routine QAQC checks and from results of spiked sample analysis tailored to this study QAQC information will be more fully discussed in subsequent sections

SAMPLE ANALYSIS METHODOLOGY

The samples as discussed above were drilled from the sides of randomly selected billets or archived cubes In order to prepare the samples for radiological analysis the solid DU Armor turningsshavings were first dissolved in nitric acid Subsequently this solution presumably a mixture of nitric acid DU Armor and any transuranics present was poured through a column containing an extraction chromatography resin designed to absorb any TRU in the liquid solution The solution eluted from the column therefore would contain only nitric acid and transuranics facilitating their detection via an alpha spectrometry system or mass spectrometry INTEC utilized the extraction chromatography method to separate TRU from DU Armor and quantitatively analyzed TRU by using alpha spectroscopy and Inductively Coupled Plasma (ICP) mass spectrometry RESL separated TRU from DU Armor by the coprecipitation method and quantitatively analyzed TRU by alpha spectrometry Both methods are well-established standard laboratory procedures

TABLE 1 COMPARISON OF LABORATORY METHODS FOR TRU IN DU ARMOR

INTEC RESL Separation of TRU from

DU Armor Ion Exchange Method Coprecipitation

Method Amount of TRU Alpha Spectrometry and ICP Mass

Spectrometry Alpha Spectrometry

As dissolution of the turnings in nitric acid could potentially be incomplete due to the presence of refractory plutonium aliquots from 16 of the 60 total samples were filtered through a 02-micron laboratory filter by INTEC The residue was separated for use in plutonium analysis (Pu-238 and Pushy239240) The dissolution residue was combined with lithium tetraborate in a process labeled lithium tetraborate fusion to facilitate the dissolution of plutonium3 At the conclusion of lithium tetraborate fusion the samples were handled as above by utilizing an alpha spectrometry system to analyze for

4

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 47: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 3

plutonium A comparison of analysis results between the samples prepared using lithium tetraborate fusion and the samples being prepared by dissolution in nitric acid was performed There was no difference in the levels of Plutonium detected It was therefore concluded that the DU Armor turnings did not contain refractory plutonium and that regular preparation of dissolution in nitric acid would be sufficient for the purposes of this study

Additionally because Pu-236 Am-243 and Pu-242 are often used as tracers in alpha spectrometry analysis RESL analyzed DU Armor samples for the presence of these isotopes RESLrsquos methodology dissolved 2 g of DU Armor samples in nitric acid and performed alpha spectrometry analysis focusing on the tracer isotopes With an MDC of approximately 02 pCig of DU RESLrsquos analysis showed no Pu-236 Am-243 or Pu-242 in DOE DU Armor samples Pu-236 Pu-242 and Am-243 were therefore deemed suitable tracers for this study

TRU IN DU RADIOCHEMISTRY RESULTS

INTEC results for Population 1 billets indicate that Am-241 Np-237 Pu-238 Pu-239240 and Tc-99 were present in DU Armor only in amounts well below the interim values set forth by the interim license amendment The lowest and highest activity concentrations for each nuclide are as follows

TABLE 2 Population 1 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig o f DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -080 ndashndash 13 44 ndashndash 55 Np-237 lt100 lt13 ndashndash NA 37 ndashndash 092 Pu-238 lt100 -003 ndashndash 006 20 ndashndash 053

Pu-239240 lt100 -12 ndashndash 19 27 ndashndash 088 Tc-99 lt500 lt73 ndashndash NA 240 ndashndash 47

NA = Not Available

Complete information on all Population 1 Samples may be found in Appendices A amp D

INTEC analysis of Population 2 billets yielded similar results excepting Tc-99 which was slightly above the interim value Values for all isotopes except Tc-99 remained well below the interim values set forth by the interim amendment Though most values for Tc-99 were well below the interim values two values samples W05199211RH and W05199411RH exceeded the interim values The lowest and highest activity concentrations are recorded below and complete analysis information for all Population 2 samples may be found in Appendices B amp D

5

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 48: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 3

TABLE 3 Population 2 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor)

Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 -17 ndashndash 28 19 ndashndash 58 Np-237 lt100 lt11 ndashndash NA 22 ndashndash NA Pu-238 lt100 001 ndashndash 001 080 ndashndash 014

Pu-239240 lt100 012 ndashndash 017 10 ndashndash 016 Tc-99 lt500 64 ndashndash NA 540 ndashndash 32 NA = Not Available

Population 3 billets yielded results similar to those from the other two populations The lowest and highest activity concentrations are recorded below and complete analysis information for all population 3 samples may be found in Appendices C amp D

TABLE 4 Population 3 Billets Highest Concentrations by Nuclide (INTEC)

Nuclide Interim Max Value (pCig of

DU Armor)

Lowest Value (pCig of DU Armor) Activity +- 1 sigma

Highest Value (pCig of DU Armor) Activity +- 1 sigma

Am-241 lt100 12 ndashndash 18 53 ndashndash 22 Np-237 lt100 12 ndashndash NA lt36 ndashndash NA Pu-238 lt100 017 ndashndash 006 086 ndashndash 023

Pu-239240 lt100 024 ndashndash 006 086 ndashndash 014 Tc-99 lt500 83 ndashndash NA 400 ndashndash 26

NA = Not Available

Collected INTEC data for all three populations including uncertainty values may be found in Appendix D

GAMMA SPECTROSCOPIC ANALYSIS OF DU BILLETS

Tc-99 a fission product was detected in approximately 50 of DU billets analyzed The US Army took a proactive approach and requested that INTEC analyze the DU Armor samples for other possible fission products by gamma spectrometric analysis An aliquot of DU Armor sample solution was counted on a High Purity high-resolution gamma spectrometric system Given the age of the DU Armor only gamma emitters with a half-life greater than 25 years were considered No gamma peaks were observed except for the uranium progeny

6

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 49: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 3

QUALITY CONTROL

INTEC INTERNAL QC

The approach to QC in this study was two-fold both internal and external checks were established to ensure accurate results Initial laboratory selection was of paramount importance and long-term QC records were consulted for each laboratory under consideration Historically INTEC has an exceptional performance record in the DOE Quality Assurance Program (QAP) a nationally known and respected QA inter-comparison program INTEC includes results from its participation in the DOE QAP as supporting information in their report entitled ldquoSMC Billets4rdquo

In performing the analysis for this study INTEC used three internal measures to assess the quality of their analyses The samples were analyzed in batches of 6-10 samples Several QC standards were included with each batch of samples The first labeled a ldquoLaboratory Control Standardrdquo or LCS contained known amounts of all TRU or fission isotopes in question (Pu-238 Pu-239240 Am-241 Np-237 and Tc-99) in pure nitric acid All solutions used for LCS tracers and matrix spikes have National Institute of Standards and Technology (NIST) traceablity The analysis of this standard allowed INTEC to gauge the percentage yield of TRU activity based on a comparison with the known activities of the isotopes in the standard

The Laboratory Control Standards (LCS) in pure nitric acid used by INTEC are shown below

TABLE 5 LABORATORY CONTROL

STANDARDS (LCS) AT INTEC Nuclide Value (pCig) Am-241 14 58 and 915 Np-237 28 Pu-238 12 53 Pu-239 67

The above LCS were analyzed with each batch of 6-10 DU Armor samples at the same time

In addition to the LCS analyzed for each batch an isotopic tracer was added to each DU Armor sample analyzed for the TRU isotopes The tracer isotopes used were Am-243 Np-239 Pu-236 and Pu-242 A certified amount of each of the above isotopes with a known activity was added to each DU Armor sample prior to chemical separation The tracers are chemically identical to the target isotopes and therefore indicate the losses incurred from the separation process By measuring the amount of tracer activity in the final sample counted and comparing to the tracer added a chemical yield can be calculated This yield factor is then used to correct the final value of the target isotope Correcting this value allowed INTEC to provide a more accurate estimate of TRU activity in DU Armor

A third QC sample analyzed with each batch of billet samples was called a ldquoUranium Matrix Controlrdquo and included a pure uranium matrix and known amounts of the following isotopes Am-241 Pu-239240 and Np-237 This control was utilized to determine the effectiveness of the ion exchange process in separating the target isotopes from the bulk uranium matrix The control yield for all isotopes generally fell within the 90-110 range indicating that the ion exchange process was quite effective

7

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 50: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 3

INTEC LABORATORY MINIMUM DETECTABLE CONCENTRATIONS

The detection limit set by the DA analysis protocol was 1 pCig of DU Armor sample The MDC levels achieved by INTEC were approximately one order of magnitude lower than the requested value Most of the MDC results were close to 01 pCig of DU Armor sample Therefore the methodologies achieved appropriate levels of sensitivity

EXTERNAL QC

As discussed above RESL participated in this study in order to provide analysis results to compare with those from INTEC RESL received 8 of the total 60 samples including 3 samples from Population 1 3 samples from Population 2 and 2 samples from Population 3 Data comparisons can be found below MDCs achieved by each lab for the compared samples can found in Table 9 Variability exists between INTEC and RESL analysis results and MDCs achieved These differences may be due to different methodologies employed and for possible sample inhomogeneities

8

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 51: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 3

TABLE 6 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 1 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim Maximum

Value lt100 lt100 lt100 lt100 lt500

Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 441 ndash 551 058 ndash 007 373 ndash 092 051 ndash 009 046 ndash 010 067 ndash 010 057 ndash 011 093 ndash 012 lt92 NA Bottom 314 ndash 135 NA 190 ndash 082 NA 045 ndash 011 NA 058 ndash 016 NA lt82 NA

66 Top 000 ndash 183 064 ndash 014 254 ndash 078 005 ndash 005 119 ndash 048 012 ndash 006 266 ndash 088 047 ndash 008 87 ndash 39 NA Bottom 173 ndash 234 NA 335 ndash 085 NA 205 ndash 053 NA 059 ndash 010 NA lt79 NA

140 Top -045 ndash 072 005 ndash 006 164 021 ndash 007 011 ndash 004 -004 ndash 004 043 ndash 009 042 ndash 008 100 ndash 44 NA Bottom 229 ndash 330 NA 167 NA 014 ndash 018 NA 039 ndash 010 NA 240 ndash 47 NA

NA = Not Available TABLE 7

Analysis of TRU in Depleted Uranium Armor Comparison of Analysis of INTEC vs RESL Population 2 Billets (Activity +- 1 sigma)

Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig)

Interim Maximum

Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

13 386 ndash 449 006 ndash 006 183 027 ndash 008 083 ndash 014 007 ndash 005 036 ndash 008 031 ndash 007 lt75 NA

16 330 ndash 386 -005 ndash 006 167 ndash 072 030 ndash 009 009 ndash 010 001 ndash 006 021 ndash 006 03 ndash 008 lt94 ndash 43 NA 17 212 ndash 305 048 ndash 010 150 035 ndash 007 005 ndash 008 004 ndash 002 041 ndash 010 051 ndash 008 120 ndash 48 NA

NA = Not Available

9

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 52: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 3

TABLE 8 Analysis of TRU in Depleted Uranium Armor

Comparison of Analysis of INTEC vs RESL Population 3 Billets (Activity +- 1 sigma) Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Value (pCig) Value (pCig) Value (pCig) Value (pCig) Value (pCig) Interim

Maximum Value

lt100 lt100 lt100 lt100 lt500

Random Sample

INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

1 324 ndash 408 029 ndash 009 214 017 ndash 007 086 ndash 023 02 ndash 007 006 ndash 019 054 ndash 009 96 NA 10 327 ndash 126 -004 ndash 004 120 029 ndash 007 025 ndash 006 012 ndash 005 056 ndash 010 059 ndash 009 400 NA

NA = Not Available

TABLE 9 Analysis of TRU in Depleted Uranium Armor

Comparison of MDC Analysis of INTEC vs RESL Nuclide Am-241 Np-237 Pu-238 Pu-239240 Tc-99

Units pCig Random Sample

Position INTEC RESL INTEC RESL INTEC RESL INTEC RESL INTEC RESL

28 Top 52 080 16 080 014 080 016 080 92 NA

Bottom 28 080 14 080 012 080 019 080 82 NA

66 Top 74 080 13 080 012 080 022 080 69 NA

Bottom 27 080 15 080 006 080 006 080 79 NA

140 Top 67 080 16 080 007 080 008 080 77 NA

Bottom 53 080 17 080 017 080 014 080 88 NA 13 40 080 18 080 011 080 009 080 75 NA 16 35 080 13 080 009 080 009 080 78 NA 17 49 080 15 080 013 080 011 080 83 NA 1 38 080 21 080 026 080 030 080 96 NA

10 23 080 12 080 008 080 007 080 66 NA NA = Not Available

10

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 53: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 3

BLIND SAMPLE ANALYSIS

Finally INTEC was asked to analyze one additional sample for quality control purposes RESL spiked a pure uranium standard with known amounts of Am-241 Np-237 Pu-238 and Pu-239 This sample was then sent to INTEC with instructions to forward the results of that analysis to the Project Officer Dr Ram Bhat CHP Dr Bhat and the RESL personnel preparing the sample were the sole possessors of the known values Upon receipt of INTECrsquos analysis Dr Bhat compared the actual activity levels in the spike to those observed experimentally by INTEC The comparison reflected favorably upon INTECrsquos analyses and results of the blind sample analysis are listed below

TABLE 10 INTEC ANALYSIS OF BLIND SAMPLE

Nuclide RESL Known Activity (pCimL)

(ndashndash 1 sigma)

INTEC Exp Activity (pCimL)

(ndashndash 1 sigma)

Yield

Am-241 113 ndash 002 109 ndash 016 96 Np-237 065 ndash 001 072 ndash 007 111 Pu-238 047 ndash 001 047 ndash 005 100 Pu-239 095 ndash 002 091 ndash 009 96

PROPAGATION OF ERRORS

Analysis of TRU in DU Armor required several critical chemicalion exchange separations and different types of counting methods to estimate the quantities of each nuclide of TRU contained in DU Armor A sample batch consisted of 8-10 DU Armor samples that were analyzed simultaneously A tracer was added to each sample to estimate the chemical yield of the tracer at the end of the sample analysis Also a LCS was run along with the sample batch to obtain the yield of the control samples In addition a TRU-spiked Uranium sample was analyzed with the sample batch to monitor the chemical separation in Uranium Matrix All yields were utilized to compute sample uncertainty values Typical hand calculated samples of the propagation of errors for RESL and INTEC are given in Appendices E and F respectively

DISCUSSION

As mentioned above the QA program established by DA for the investigation of TRU in DU Armor samples was quite extensive Apart from the usual quality control of instrumentation methodology standard operating procedures duplicate analysis and sample chains of custody DA has employed two separate laboratories in this investigation of TRU in DU Armor for quality assurance purposes Both laboratories have an excellent record of participation in the DOE-monitored QAP and MAPEP programs In this study however some apparent variabilities exist between INTEC and RESL analysis results Finally the litmus test for QA in this study was the excellent performance of INTEC in blind spiked analysis of TRU in uranium as displayed in Table 10

11

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 54: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 3

The population 1 billets have the following dimensions 16rdquo x 18rdquo x 2rdquo Two samples were taken from each random billet as shown in Appendix A to check the homogenous distribution of TRU and Tc-99 in DU Armor TRU and Tc-99 in DU Armor of the top and bottom sections of these 10 random billets agree well within experimental error This agreement indicates that TRU and Tc-99 are distributed uniformly in the DU Armor billets Hence similar comparison of top and bottom sections of the DU Armor billets is not carried out in population 2 and 3 DU Armor billets

The results of this analysis indicate to a reasonable degree of certainty that Am-241 Np-237 Pu-238 and Pu-239240 content in DU Armor is minimal The TRU levels that have been detected are similar across all three populations

The results of this analysis indicate that Tc-99 content in DU Armor slightly exceeded the interim value for two samples out of 60 samples analyzed

TRU IN CONTEXT

An assessment was undertaken to determine the extent of increased radiological health and safety risk associated with trace amounts of TRU in the DU Armor As TRU and Tc-99 emit particulate radiation (alpha beta and low energy photons) and the DU Armor is encased in steel the presence of trace amounts of TRU in DU Armor should not result in a measurable difference in external dose This conclusion is consistent with a previous DOE safety analysis review5

In order to assess internal dose the percent increase in risk (in fractions of Annual Limit on Intake (ALIs)) due to the interim maximum value of TRU (100 pCi of Am-241 100 pCi of Npshy237 100 pCi of Pu-238 100 pCi of Pu-239240) and Tc-99 (500 pCi) for 1 g of inhaled DU Armor was calculated and compared to the ALI calculated for 1 g of inhaled DU Armor

The ALIs used in the following equations are listed below in Table 11(REF 10CFR20 App B)

TABLE 11 ALIs OF TRU Tc-99 AND DU Nuclide ALI (pCi) Class Am-241 6 X 103 W Np-327 4 X 103 W Pu-238 7 X 103 W

Pu-239240 6 X 103 W Tc-99 7X108 W

DU (U-234 U-235 U-236 U-238)

4X104 Y

Percent ALI TRU and Tc-99

[100 pCi Am-241ALI Am-241] + [100 pCi Np-237ALI Np-237]+ [100 pCi Pu-238 ALI Pushy238] + [100 pCi Pu-239240ALI Pu-239240] + [500 pCi Tc-99ALI Tc-99]

= 0073 ALI

12

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 55: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 3

The specific activity of 1 g of DU is 3775X105 pCi

3775X105 pCi DU ALI (4X104 pCi) = 94 ALIs

The Ratio of the ALIs of TRU Tc-99 and DU

0073 ALIs94 ALIs = 08 (or less than a 1 increase in risk as represented by the ALI)

In the above equation if 10000 pCi of Tc-99 is used in place of 500 pCi of Tc-99 the total percentage of TRU and Tc-99 ALI still equals 0073 There is no change in the percent ALI because the ALI of Tc-99 is 7X108 pCi

The corresponding mass concentration for interim maximum TRUs and Tc-99 are given in Table 12

TABLE 12 ACTIVITY AND MASS CONCENTRATION OF INTERIM MAXIMUM

TRUFP CONTAMINANTS Nuclide Specific

Activity (pCig)

Activity Concentration

(pCig of DU Armor)

Ratio Contaminant

DU Armor (mgg)

Am-241 34X1012 100 29X10-8

Np-237 71X108 100 14X10-4

Pu-238 17X1013 100 58X10-9

Pu-239240 62X1010 100 16X10-6

Tc-99 17X1010 500 30X10-5

Specific Activity of Pu-239 only

CONCLUSIONS AND RECOMMENDATIONS

As evidenced by the risk analysis approach employed above the interim values of 100 pCig of each TRU and 500 pCig Tc-99 result in an increase of only 08 to the overall occupational risk as measured by ALI None of the TRU values identified by analysis of 60 DU Armor samples from three different populations of billets approached 100 pCig The maximum TRU value was 19 ndash 58 pCig of Am-241 in population 2 well below the interim value of 100 pCig However two samples out of 60 DU Armor billets slightly exceeded the interim value of 500 pCig for Tc-99 But as evidence by the percentage ALI TRU and Tc-99 equation shows even an increase to 10000 pCig of Tc-99 will not increase the overall occupational risk above 08 Even though two samples out of 60 DU Armor billets had Tc-99 values greater than 500 pCig (510 ndash 30 pCig and 540 ndash32 pCig) the overall occupational risk (as represented by the ALI) still will not exceed 08

13

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 56: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 3

It is also important to underscore that the radiation protection program currently in place in support of the DU Armor program also adequately protects against these minute quantities of TRU and Tc-99

Therefore the presence of these trace radionuclides in DU Armor is safe

REFERENCES

1 Idaho National Engineering and Environmental Laboratory Lmitco Idaho Falls ID 83415 ldquoAbbreviate Sampling and Analysis Plan for SMC Depleted Uraniumrdquo WG5-051-99 September 30 1999

2 ldquoQuality Assurance Project Plan for the Analytical Laboratories Department Radioanalytical Sectionrdquo December 31 1997 Lockheed Martin Idaho Technologies Company

3 CW Sill and DS Sill ldquoSample Dissolutionrdquo Radioactivity and Radiochemistry 6(2) 8 (1995)

4 ldquoSMC Billetsrdquo INEELINT-99-01228 INTEC Radiochemistry December 151999

5 ldquoAnalysis of the Transuranic Contamination of the Depleted Uranium Draftrdquo Internal Draft Report dated September 9 1999

14

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 57: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 3

Appendix A Analysis of TRU in Depleted Uranium Armor

INTEC Samples of Population 1 Billets

Random Position Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500

10

Top W05199011RH 20 lt 14 -003 035 110 Bottom W05199021RH 25 lt 13 010 023 130

28

Top W05199031RH 44 37 050 057 lt 92 Bottom W05199041RH 31 19 050 058 lt 82

47

Top W05199051RH 055 lt 15 003 -002 lt 78 Bottom W05199061RH 16 lt 16 006 006 lt 73

66

Top W05199071RH 000 25 120 27 87 Bottom W05199081RH 17 34 200 059 lt 79

84

Top W05199091RH 16 32 032 027 lt 92 Bottom W05199101RH 29 14 033 066 150

103

Top W05199111RH 36 lt 16 019 028 lt 80 Bottom W05199121RH 42 17 009 034 lt 79

122

Top W05199131RH -080 18 012 031 lt 74 Bottom W05199141RH 24 lt 15 003 -12 lt 82

140

Top W05199151RH -045 lt 16 011 043 100 Bottom W05199161RH 23 lt 17 014 039 240

159

Top W05199171RH 24 lt 15 026 037 lt 84 Bottom W05199181RH 15 lt 16 014 026 lt 87

170

Top W05199191RH 000 lt 15 011 014 lt 83 Bottom W05199201RH 21 lt 16 023 018 lt 81

15

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 58: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 3

Appendix B Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 2 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99 Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199211RH 49 lt 15 062 10 510 W05199221RH 12 lt 18 023 034 lt 94 W05199231RH 19 lt 20 016 021 lt 95 W05199241RH 53 lt 18 022 020 lt 83 W05199251RH 34 lt 13 014 043 lt 64 W05199261RH 23 lt 19 016 021 lt 89 W05199271RH 36 lt 20 036 055 lt 93 W05199281RH 10 lt 14 018 012 110 W05199291RH -020 lt 18 021 016 lt 84 W05199301RH -17 lt 14 023 041 99 W05199311RH -16 lt 16 017 019 lt 88 W05199321RH 31 lt 16 053 047 lt 85 W05199331RH 39 lt 18 083 036 lt 75 W05199341RH 40 lt 17 020 049 lt 90 W05199351RH 20 lt 16 007 058 240 W05199361RH 33 17 009 021 95 W05199371RH 21 lt 15 005 041 120 W05199381RH 15 lt 20 012 043 lt 96 W05199391RH 25 lt 18 021 069 170 W05199401RH 34 lt 22 007 033 230 W05199411RH 31 lt 15 021 059 540 W05199421RH 44 lt 14 009 044 160 W05199431RH 10 lt 15 001 037 lt 82 W05199441RH 039 18 008 017 190 W05199451RH 15 lt 13 009 023 330 W05199461RH 26 lt 12 017 063 280 W05199471RH 37 lt 14 019 043 270 W05199481RH 25 lt 11 018 055 430 W05199491RH 50 16 009 030 270 W05199501RH 19 lt 14 014 038 280

16

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 59: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 3

Appendix C Analysis of TRU in Depleted Uranium Armor INTEC Samples of Population 3 Billets

Sample ID Nuclide

Am-241 Np-237 Pu-238 Pu-239240 Tc-99Units Value (pCig of DU Armor)

Interim Values lt100 lt100 lt100 lt100 lt500 W05199511RH 32 lt 21 086 060 lt 96 W05199521RH 12 lt 18 044 086 lt 89 W05199531RH 36 lt 19 041 053 140 W05199541RH 53 lt 31 048 060 330 W05199551RH 36 lt 18 045 074 220 W05199561RH 39 lt 36 017 029 lt 93 W05199571RH 19 lt 21 028 024 180 W05199581RH 25 lt 15 029 059 lt 83 W05199591RH 28 lt 15 054 072 360 W05199601RH 33 lt 12 025 056 400

17

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 60: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 3

A p p e n d i x D I N T E C D U A r m o r S a m p l e A n a l y s i s a n d U n c e r t a i n t i e s ( + - 1 s i g m a ) N u c l i d e A m - 2 4 1 N p - 2 3 7 P u - 2 3 8 P u - 2 3 9 2 4 0 T c - 9 9

Uni ts p C i g o f D U A r m o r

I n t e r i m V a l u e s lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 1 0 0 lt 5 0 0

S a m p l e I D V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c V a l u e U n c P o p W 0 5 1 9 9 0 1 1 R H 2 0 2 8 lt 1 4 -0 03 0 06 0 35 0 07 1 1 0 3 9 1 W 0 5 1 9 9 0 2 1 R H 2 5 1 1 lt 1 3 0 10 0 11 0 23 0 07 1 3 0 3 8 1 W 0 5 1 9 9 0 3 1 R H 4 4 5 5 3 7 0 92 0 50 0 10 0 57 0 11 lt 9 2 1 W 0 5 1 9 9 0 4 1 R H 3 1 1 4 1 9 0 82 0 50 0 11 0 58 0 16 lt 8 2 1 W 0 5 1 9 9 0 5 1 R H 0 55 0 85 lt 1 5 0 03 0 04 -0 02 0 04 lt 7 8 1 W 0 5 1 9 9 0 6 1 R H 1 6 2 3 lt 1 6 0 06 0 07 0 06 0 07 lt 7 3 1 W 0 5 1 9 9 0 7 1 R H 0 00 1 8 2 5 0 78 1 2 0 48 2 7 0 88 8 7 3 9 1 W 0 5 1 9 9 0 8 1 R H 1 7 2 3 3 4 0 85 2 0 0 53 0 59 0 10 lt 7 9 1 W 0 5 1 9 9 0 9 1 R H 1 6 2 4 3 2 1 0 0 32 0 12 0 27 0 36 lt 9 2 1 W 0 5 1 9 9 1 0 1 R H 2 9 3 6 1 4 0 68 0 33 0 07 0 66 0 11 1 5 0 4 2 1 W 0 5 1 9 9 1 1 1 R H 3 6 1 4 lt 1 6 0 19 0 06 0 28 0 07 lt 8 0 1 W 0 5 1 9 9 1 2 1 R H 4 2 1 7 1 7 0 82 0 09 0 14 0 34 0 12 lt 7 9 1 W 0 5 1 9 9 1 3 1 R H -0 80 1 3 1 8 0 76 0 12 0 04 0 31 0 08 lt 7 4 1 W 0 5 1 9 9 1 4 1 R H 2 4 2 9 lt 1 5 0 03 0 05 -1 2 1 9 lt 8 2 1 W 0 5 1 9 9 1 5 1 R H -0 45 0 72 lt 1 6 0 11 0 04 0 43 0 09 1 0 0 4 4 1 W 0 5 1 9 9 1 6 1 R H 2 3 3 3 lt 1 7 0 14 0 18 0 39 0 10 2 4 0 4 7 1 W 0 5 1 9 9 1 7 1 R H 2 4 3 2 lt 1 5 0 26 0 08 0 37 0 09 lt 8 4 1 W 0 5 1 9 9 1 8 1 R H 1 5 2 0 lt 1 6 0 14 0 19 0 26 0 32 lt 8 7 1 W 0 5 1 9 9 1 9 1 R H 0 00 1 4 lt 1 5 0 11 0 12 0 14 0 16 lt 8 3 1 W 0 5 1 9 9 2 0 1 R H 2 1 3 0 lt 1 6 0 23 0 07 0 18 0 07 lt 8 1 1 W 0 5 1 9 9 2 1 1 R H 4 9 1 9 lt 1 5 0 62 0 11 1 0 0 16 5 1 0 3 0 2 W 0 5 1 9 9 2 2 1 R H 12 3 2 9 lt 1 8 0 23 0 09 0 34 0 10 lt 9 4 2 W 0 5 1 9 9 2 3 1 R H 19 2 5 8 lt 2 0 0 16 0 21 0 21 0 09 lt 9 5 2 W 0 5 1 9 9 2 4 1 R H 5 3 2 2 lt 1 8 0 22 0 06 0 20 0 06 lt 8 3 2 W 0 5 1 9 9 2 5 1 R H 3 4 1 4 lt 1 3 0 14 0 05 0 43 0 09 lt 6 4 2 W 0 5 1 9 9 2 6 1 R H 2 3 3 3 lt 1 9 0 16 0 19 0 21 0 07 lt 8 9 2 W 0 5 1 9 9 2 7 1 R H 3 6 5 2 lt 2 0 0 36 0 09 0 55 0 12 lt 9 3 2 W 0 5 1 9 9 2 8 1 R H 1 0 1 4 lt 1 4 0 18 0 06 0 12 0 17 1 1 0 4 1 2 W 0 5 1 9 9 2 9 1 R H -0 20 0 31 lt 1 8 0 21 0 07 0 16 0 06 lt 8 4 2 W 0 5 1 9 9 3 0 1 R H -1 7 2 8 lt 1 4 0 23 0 08 0 41 0 10 9 9 4 4 2 W 0 5 1 9 9 3 1 1 R H -1 6 2 5 lt 1 6 0 17 0 22 0 19 0 23 lt 8 8 2 W 0 5 1 9 9 3 2 1 R H 3 1 3 8 lt 1 6 0 53 0 11 0 47 0 11 lt 8 5 2 W 0 5 1 9 9 3 3 1 R H 3 9 4 5 lt 1 8 0 83 0 14 0 36 0 08 lt 7 5 2 W 0 5 1 9 9 3 4 1 R H 4 0 5 5 lt 1 7 0 20 0 08 0 49 0 11 lt 9 0 2 W 0 5 1 9 9 3 5 1 R H 2 0 2 8 lt 1 6 0 07 0 10 0 58 0 12 2 4 0 4 5 2 W 0 5 1 9 9 3 6 1 R H 3 3 3 9 1 7 0 72 0 09 0 10 0 21 0 06 9 5 4 4 2 W 0 5 1 9 9 3 7 1 R H 2 1 3 1 lt 1 5 0 05 0 08 0 41 0 10 1 2 0 4 8 2 W 0 5 1 9 9 3 8 1 R H 1 5 2 3 lt 2 0 0 12 0 06 0 43 0 11 lt 9 6 2 W 0 5 1 9 9 3 9 1 R H 2 5 3 7 lt 1 8 0 21 0 07 0 69 0 14 1 7 0 4 8 2 W 0 5 1 9 9 4 0 1 R H 3 4 4 6 lt 2 2 0 07 0 03 0 33 0 08 2 3 0 4 8 2 W 0 5 1 9 9 4 1 1 R H 3 1 3 6 lt 1 5 0 21 0 06 0 59 0 12 5 4 0 3 2 2 W 0 5 1 9 9 4 2 1 R H 4 4 4 9 lt 1 4 0 09 0 11 0 44 0 10 1 6 0 4 9 2 W 0 5 1 9 9 4 3 1 R H 1 0 1 5 lt 1 5 0 01 0 01 0 37 0 08 lt 8 2 2 W 0 5 1 9 9 4 4 1 R H 0 39 0 60 1 8 0 82 0 08 0 11 0 17 0 06 1 9 0 4 8 2 W 0 5 1 9 9 4 5 1 R H 1 5 1 9 lt 1 3 0 09 0 04 0 23 0 06 3 3 0 2 9 2 W 0 5 1 9 9 4 6 1 R H 2 6 3 1 lt 1 2 0 17 0 07 0 63 0 13 2 8 0 3 1 2 W 0 5 1 9 9 4 7 1 R H 3 7 1 3 lt 1 4 0 19 0 05 0 43 0 09 2 7 0 2 8 2 W 0 5 1 9 9 4 8 1 R H 2 5 3 3 lt 1 1 0 18 0 05 0 55 0 10 4 3 0 2 7 2 W 0 5 1 9 9 4 9 1 R H 5 0 1 7 1 6 0 80 0 09 0 11 0 30 0 08 2 7 0 3 5 2 W 0 5 1 9 9 5 0 1 R H 1 9 2 7 lt 1 4 0 14 0 05 0 38 0 08 2 8 0 3 5 2 W 0 5 1 9 9 5 1 1 R H 3 2 4 1 lt 2 1 0 86 0 23 0 60 0 19 lt 9 6 3 W 0 5 1 9 9 5 2 1 R H 1 2 1 8 lt 1 8 0 44 0 10 0 86 0 14 lt 8 9 3 W 0 5 1 9 9 5 3 1 R H 3 6 4 5 lt 1 9 0 41 0 09 0 53 0 11 1 4 0 5 3 3 W 0 5 1 9 9 5 4 1 R H 5 3 2 2 lt 3 1 0 48 0 10 0 60 0 12 3 3 0 3 8 3 W 0 5 1 9 9 5 5 1 R H 3 6 4 2 lt 1 8 0 45 0 11 0 74 0 13 2 2 0 4 5 3 W 0 5 1 9 9 5 6 1 R H 3 9 1 6 lt 3 6 0 17 0 06 0 29 0 08 lt 9 3 3 W 0 5 1 9 9 5 7 1 R H 1 9 2 7 lt 2 1 0 28 0 07 0 24 0 06 1 8 0 5 0 3 W 0 5 1 9 9 5 8 1 R H 2 5 3 5 lt 1 5 0 29 0 07 0 59 0 11 lt 8 3 3 W 0 5 1 9 9 5 9 1 R H 2 8 1 1 lt 1 5 0 54 0 10 0 72 0 12 3 6 0 3 2 3 W 0 5 1 9 9 6 0 1 R H 3 3 1 3 lt 1 2 0 25 0 06 0 56 0 10 4 0 0 2 6 3

18

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 61: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 3

APPENDIX E FORMULAS AND SAMPLE CALCULATIONS

RESL DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR PU-239 USING RESL DATA

EQUATION 1

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

EQUATION 2

2 cts239 + bkgd239

2 ctstr + bkgdtr

2 smL _ tracer

2 sact _ tracer

2 sg

s = [ pCi g ]x + + + + cts239 - bkgd239 ctstr - bkgdtr mL _ tracer act_ tracer gŁ ł Ł ł Ł ł Ł ł Ł ł

Where

cts239 = total counts of Pu-239

bkgd239 = total background of Pu-239

ctstr = total counts of tracer (Pu-242 of Pu-236)

bkgdtr = total background counts of tracer (Pu-242 or Pu-236)

mL_tracer = amount of tracer added (g or mL)

smL_tracer = standard deviation of the amount of tracer added (g or mL)

act_tracer = activity of the tracer (pCimL or pCig)

sact_tracer = standard deviation of the activity of the tracer (pCimL or pCig)

g = weight of sample taken for analysis

sg = standard deviation of the sample weight

Appendix E-1

19

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 62: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 3

EXAMPLE CALCULATIONAND UNCERTAINTY ANALYSIS FOR PU-239 FROM RESL DATA

1 Using the data from Encl 1 with data from Sample Number W05199032RH

Where

cts239 = 71 counts bkgd239 = 2 counts

ctstr = 1087 counts bkgdtr = 3 counts

mL_tracer = 05 mL smL_tracer = ndash0002 mL

act_tracer = 388 pCimL sact_tracer = ndash004 pCimL

g = 0133 g sg = ndash002 g

2 Placing these numbers into Equation 1 determines the activity of the Pu-239

(71counts - 2counts) x (05mL ndash 0002 mL x 388 pCi mL ndash 004 pCi mL) = pCi g

(1087 counts - 3counts) x 0133 g ndash 0002 g

(cts239 - bkgd239) x(mL _tracer ndash smL _tracerxact_ tracer ndashsact _tracer) = pCi g

(ctstr- bkgdtr)xg ndashsg

(69counts) x(0500mLx388 pCi mL) = pCi g

(1084counts)x0133g

Activity of Pu-239 = 093 pCig

3 Using the Activity from Equation 1 and placing it into Equation 2 the Uncertainty is determined for Pu-239

cts239 + bkgd239 2

ctstr+ bkgdtr2

smL_ tracer 2

sact _tracer 2

sg s = [ pCi g]x + + + +

Ł cts239 - bkgd239 ł Ł ctstr- bkgdtr ł Ł mL_ tracer ł Ł act_ tracer ł Ł g ł

Appendix E-2

20

2

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 63: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

2

Attachment 3

2 2 2 71 + 3counts 1087 + 3counts 0002 mL 004 pCi mL 2

0002 g s = [093 pCi g ]x + + + +

71 - 3counts 1087 - 3counts Ł 05mL ł 388 pCi mL 0133 gŁ ł Ł ł Ł ł Ł ł

22 2 2 2

854 3302 0002 004 0002 s = [093pCi g]x + + + + Ł 69 ł Ł 1084 ł Ł 05 ł Ł 388 ł Ł 0133ł

s = [093pCi g]x (00153) + (000093) + (0000016) + (0000106) + (0000226)

s = [093 pCi g]x (00165781)

Standard Deviation of Pu-239 = 012 pCig

Appendix E-3

21

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 64: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 3

APPENDIX F FORMULAS AND SAMPLE CALCULATIONS

INTEC DATA

FORMULA FOR SAMPLE CALCULATION AND UNCERTAINTY ANALYSIS FOR Pushy239 USING INTEC DATA

Sample Number W05199271RH (INTEC Number 9CE33) Isotope Pu-239 Tracer Pu-236

Decay Correction of Tracer

Pu-236 = 6132 BqmL on 42398 Dt =153 yrs

Pu-236 (BqmL) on 11299 is given by

ln 2 x (153y ) 287yPu - 236 = 6132e = 42375Bq mL

tracer added = 01 mL 042375 Bq tracer added

Calculation of Sample Results

(net _ cnts _ sple)x(dil )Pu - 239 Bq g =

(eff )x(t )x( y)x( BR )Pu-239

Where net_cnts_sple = gross count Pu-239 ndash Bkg counts of Pu-239 eff = absolute detector efficiency t = count time in seconds BRPu-239 = branching ratio Pu-239 y = tracer yeild dil = dilution factor

(net _ tracer _ counts)y =

(eff )x(t)x(BR ) x(tracer _ added _ Bq)Pu-236

(net _ cnts _ sple)x(eff ) x(t) x(BR )x(tracer _ added _ Bq )x(dil)Pu-236 Pu - 239 (Bq g ) = (eff ) x(t) x(BR )x(net _ tracer _ cts)Pu-239

Appendix F-1

22

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 65: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 3

(net _ cnts_ Pu - 239)x(100)x(042375Bq)(dil)=

(0999) x(net _ cnts_ Pu - 236_ tracer)

For sample W05199271RH

50mL -1dil = = 274g(996mL)x(1832g)

net_cnts_Pu-239 = 544 counts

net_cnts_Pu-236_tracer = 30747 counts

(544counts) x(100)x(042375 Bq) x(274 g -1) 2 Pu - 239 (Bq g ) = = 205 X 10 - Bq g(0999)x(3074 7counts)

Uncertainty Calculations

UTotal = Urand + Usys = Total Relative Uncertainty

net _ cnts _ sple

1 2 2 2 2 2 2Usys = (Uspk + UI + D U HL + UL )

Where UI = relative uncertainty in the spike peak branching ratio (intensity) D = natural logarithm of decay correction factor UHL = relative uncertainty of nuclide half-life Uspk = total relative uncertainty of tracer UL = relative uncertainty due to laboratory sample prep

Uspk = Urand_tracer + Usys tracer

Where

total _ cnts_tracerU = rand _ tracer net _ cnts_ tracer

Appendix F-2

splecntstotalUrand

__ =

23

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448
Page 66: de maximis, inc., inc.While it is in fact possible that some disposed uranium could resist the hot nitric digestion, it would also certainly resist leaching by rainwater infiltration

Attachment 3

1 2 2 2 2 2U = (U +U + D U +U ) 2

sys _ tracer eff I HL SA

Where Ueff = relative uncertainty of detector efficiency calibration UI = relative uncertainty of tracer BR D = natural log of decay correction factor UHL = relative uncertainty of tracer nuclide half-life USA = relative uncertainty of the tracer activity value

For sample W05199271RH(9CE33)

2 2 2 2 -22Usys tracer = [(001443374) + (00012) + (000281) + (00391) ]1

= 418X10 _

3079 -2U = = 180X10rand _ tracer 30747

-2 -2 -2Uspk = 180 X10 + 418 X10 = 598 X10

58 -= = 140X10 1Urand 544

2 2 2 2) -2Usys = [(00598) + (001001) + (000066) + (002) ]12 = 638X10

UTotal = 140X10-1 + 638X10-2 = 204X10-1

Total Uncertainty = (Total relative uncertainty)(Sample result) =(204X10-1)(205X10-2 Bqg) = 418X10-3 Bqg

Activity of Pu-239 = 205X10-2 ndash 418X10-3 Bqg (Hand Cal)

Activity of Pu-239 = 2052X10-2 ndash 4269X10-3 Bqg (Computer Cal)

Appendix F-3

24

  • Response to CREW Radiological Review 4 3 06pdf
    • Attachmentspdf
      • Attach 2 - GL-MA-E-009pdf
        • Attach 2 - GL-MA-E-009pdf
          • STANDARD OPERATING PROCEDURE
          • FOR
          • 10Standard Operating Procedure for the Acid Digestion of Sediments Sludges And Soils
          • 20Purpose
          • 30Discussion
          • 40Definitions
          • 50Procedures
          • 60Safety Health and Environmental Hazards
          • 70Records Management
          • 81laboratory waste
          • 90References
            • APPENDIX 1 SAMPLE PREP LOGBOOK
              • Attach 2 - GL-RAD-A-015pdf
                • 10STANDARD OPERATING PROCEDURE FOR DIGESTION FOR SOIL
                • 20METHOD OBJECTIVE PURPOSE CODE AND SUMMARY
                • 30METHOD APPLICABILITY
                • 40DEFINITIONS
                • 50METHOD VARIATIONS
                • 60SAFETY PRECAUTIONS AND WARNINGS
                • 70INTERFERENCES
                • 80APPARATUS MATERIALS REAGENTS EQUIPMENT AND INSTRUMENTATION
                • 90SAMPLE HANDLING AND PRESERVATION
                • 100SAMPLE PREPARATION
                • 110PREPARATION OF STANDARD SOLUTIONS AND QUALITY CONTROL STANDARDS
                • 120INSTRUMENT CALIBRATION AND PERFORMANCE
                • 130ANALYSIS PROCEDURES AND INSTRUMENT OPERATION
                • 140EQUIPMENT AND INSTRUMENT MAINTENANCE
                • 150DATA RECORDING CALCULATION AND REDUCTION METHODS
                • 160QUALITY CONTROL REQUIREMENTS
                • 170DATA REVIEW APPROVAL AND TRANSMITTAL
                • 180RECORDS MANAGEMENT
                • 190LABORATORY WASTE HANDLING AND WASTE DISPOSAL
                • 200REFERENCES
                  • Appendix 1
                  • Appendix 2
                      1. barcode 568448
                      2. barcodetext SDMS Doc ID 568448