pseg early site permit application, response to request ...christine t. neely director regulatory...

Nuclear Development244 Chestnut Street, Salem, NJ 08079

2 PSEGPower LLC

ND-2012-0071October 19, 2012

U.S. Nuclear Regulatory CommissionATTN: Document Control DeskWashington, DC 20555-0001

Subject: PSEG Early Site Permit ApplicationDocket No. 52-043Response to Request for Additional Information, No. Env-12, ESP EIS6.0 - Fuel Cycle, Transportation, and Decommissioning

References: 1) PSEG Power, LLC Letter No. ND-2012-0031 to USNRC, Submittal ofRevision 1 of the Early Site Permit Application for the PSEG Site,dated May 21, 2012

2) Env-12, Review Section: ESP EIS 6.0 - Fuel Cycle, Transportation,and Decommissioning, dated September 20, 2012 (eRAI 6740)

The purpose of this letter is to respond to the request for additional information (RAI)identified in Reference 2 above. This RAI addresses Question Nos. ESP EIS 6.0-1through ESP EIS 6.0-10 for the Environmental Report (ER), as submitted in Part 3 ofthe PSEG Site Early Site Permit Application, Revision 1.

Enclosure 1 provides our response for RAI No. Env-12, Question Nos. ESP EIS 6.0-1through ESP EIS 6.0-10 (rTR-01, rTR-05a, rTR-05b, rTR-06, rTR-08, rTR-09a, rTR-11,rTR-12, rTR-13, and rTR-14).

Enclosure 2 includes the revisions to the ER resulting from our responses to RAI No.Env-12.

If any additional information is needed, please contact David Robillard, PSEG NuclearDevelopment Licensing Engineer, at (856) 339-7914.

J)19 -

U. S. Nuclear RegulatoryCommission

2 10/19/12

I declare under penalty of perjury that the foregoing is true and correct. Executed onthe 19th day of October, 2012.

Sincerely,

Christine T. NeelyDirector Regulatory AffairsPSEG Nuclear, LLC

Enclosure 1:

Enclosure 2:

Response to NRC Request for Additional Information, RAI No. Env-12,Question Nos. ESP EIS 6.0-1 through ESP EIS 6.0-10 (rTR-01, rTR-05a,rTR-05b, rTR-06, rTR-08, rTR-09a, rTR-1 1, rTR-1 2, rTR-1 3, and rTR-14), Review Section: ESP EIS 6.0 - Fuel Cycle, Transportation, andDecommissioningProposed Revisions, Part 3 - Environmental Report (ER)

cc: USNRC Project Manager, Division of New Reactor Licensing, PSEG Site(w/enclosures)USNRC Environmental Project Manager, Division of New Reactor Licensing(w/enclosures)USNRC Region I, Regional Administrator (w/enclosures)Oak Ridge National Laboratory

PSEG Letter ND-2012-0071, dated October 19, 2012

ENCLOSURE 1

RESPONSE to RAI No. Env-12

QUESTION Nos.ESP EIS 6.0-1 (rTR-01)ESP EIS 6.0-2 (rTR-05a)ESP EIS 6.0-3 (rTR-05b)ESP EIS 6.0-4 (rTR-06)ESP EIS 6.0-5 (rTR-08)ESP EIS 6.0-6 (rTR-09a)ESP EIS 6.0-7 (rTR-11)ESP EIS 6.0-8 (rTR-12)ESP EIS 6.0-9 (rTR-13)

ESP EIS 6.0-10 (rTR-14)

Response to RAI No. Env-12, Question ESP EIS 6.0-1

In Reference 2, the NRC staff asked PSEG for information regarding Transportation ofRadioactive Materials, as described in Subsection 5.7.2 of the Environmental Report.The specific request was:

rTR-01: Provide an evaluation of the heat load for an irradiated fuel shipment.

Table S-4 of 10 CFR 51.52(c) contains a heat load criterion for irradiated fuelshipments (Heat (per irradiated fuel cask in transit), 250, 000 B TU/hr.).

PSEG Response to NRC RAI:

The WASH-1238 analysis that determined the 10 CFR 51.52 Table S-4 value of theheat load criterion per irradiated fuel cask in transit assumes a cooling period of 150days prior to shipment of any irradiated fuel assemblies. Currently, there is a five yearminimum cooling period as specified by 10 CFR 961.11. Therefore, the spent fuel heatload for the reactor technologies being considered is less than the value of 250,000Btu/hr given in 10 CFR 51.52 Table S-4.

The heat load of the US-APWR spent fuel assemblies is 26,888 Btu/hr as stated in theComanche Peak Unit 3 & 4 COLA Environmental Report (ER) Section 3.8.1.11. Theheat load of the U.S. EPR is 18,600 Btu/hr as stated in the Calvert Cliffs Unit 3 COLAER Section 5.11.2. The heat load of the AP1000 is 8,330 Btu/hr as stated in the V.C.Summer Unit 2 & 3 COLA ER Section 5.11.1.8. The ABWR heat load is expected to beless than the US-APWR heat load since the ABWR has a lower burnup. Therefore, theUS-APWR is the bounding value for spent fuel heat load, and it is less than the value of250,000 Btu/hr given in 10 CFR 51.52 Table S-4.

New Subsection 5.7.2.1.12 will be included in a future revision of the ER that specifiesthe heat load per irradiated fuel cask in transit. Table 5.7-4 will be updated to includethe heat load for the reactor technologies being considered.

Associated PSEG Site ESP Application Revisions:

New ER Subsection 5.7.2.1.12 will be included and Table 5.7-4 will be updated asspecified in Enclosure 2 of this document.

Enclosure 2 includes a markup of the proposed ER revision.

Enclosure 1 Page 1


In Reference 2, the NRC staff asked PSEG for information regarding TransportationAccidents, as described in Subsection 7.4 of the Environmental Report. The specificrequest was:

rTR-05a: Provide transportation accident impacts that account for under-reporting in the state-level accident, fatality, and injury rates in ER Reference 7.4-1.

The reactors proposed do not meet all requirements of 10 CFR 51.52(a);therefore, this information is needed to determine compliance with 10 CFR51.52(b), which requires that: "...the statement shall contain a full description anddetailed analysis of the environmental effects of transportation of fuel and wastesto and from the reactor, including values for the environmental impact undernormal conditions of transport and for the environmental risk from accidents intransport."

Note: Under-reporting of the state-level accident, injury, and fatality rates is

discussed in the Comanche Peak EIS (NUREG-1943), Section 6.2.1.3, pp. 6-21.


The state-level accident, fatality, and injury rates used in Revision 1 of theEnvironmental Report (ER) are taken from ANUESD/TM-150 State-Level AccidentRates of Surface Freight Transportation: A Reexamination (ER Reference 7.4-1). TheUniversity of Michigan Transportation Research Institute (UMTRI) evaluated the samedata used in ER Reference 7.4-1 and determined that the state-level accident, fatality,and injury rates are under-reported. For the period 1994 to 1996, the accident, injury,and fatal involvements are under-reported by 39 percent, 16 percent, and 36 percent,respectively.

To account for the impact of under-reporting on the transportation of new fuel, spentfuel, and radwaste shipment transportation accidents, the accident, injury, and fatalityconsequences reported in the ER are scaled up by 1.64, 1.20, and 1.57, respectively.10 CFR 51.52, Table S-4, specifies consequence limits for fatalities and injuries only.This approach is consistent with NUREG-1943, Section 6.2.1.3, page 6-21. Table ESPEIS 6.0-2-1 shown below provides the calculated and scaled values used to evaluatethe transportation accident consequences.

Enclosure 1 Page 2

Table ESP EIS 6.0-2-1: Scaled Non-radiological Transportation Consequences

Consequence Calculated Value Scaling Factor Scaled ValueFatalities per Shipment

New Fuel 4.67E-05 1.57 7.33E-05Spent Fuel 4.68E-05 1.57 7.35E-05Radwaste 1.33E-05 1.57 2.09E-05

Injuries per ShipmentNew Fuel 1.02E-03 1.20 1.22E-03Spent Fuel 1.02E-03 1.20 1.22E-03Radwaste 3.54E-04 1.20 4.25E-04

In Tables 7.4-6 and 7.4-7 of the ER, the entries in the columns labeled 'Fatalities perShipment' and 'Injuries per Shipment' will be changed from the calculated values to thescaled values shown in Table ESP EIS 6.0-2-1. The bounding consequences aresummarized in ER Table 7.4-8 and will be updated to reflect the new values.


Subsection 7.4.1.2, Transportation of Irradiated Fuel, will be updated in a future revisionof the ER to address the under-reported accident, injury, and fatality rates. The updatewill indicate that the impacts due to transportation accidents determined by RADTRANare scaled to account for the under-reporting.

The references will be updated to include the Motor Carrier Management InformationSystem and the UMTRI 2003-6 Report. These references provide the basis foraccounting for under-reporting.

Table 7.4-6 and Table 7.4-7 of the ER will be updated to reflect the scaled results.Table 7.4-8 will be updated to reflect the new bounding reactor values.

NOTE: The last paragraph of Subsection 7.4.1.2 and Table 7.4-5 will be updated toreflect a correction in the irradiated fuel (and the total) radiological accident analysisresults.


Enclosure 1 Page 3


In Reference 2, the NRC staff asked PSEG for information regarding TransportationAccidents, as described in Section 7.4 of the Environmental Report. The specificrequest was:

rTR-05b: Provide a corrected version of ER Table 7.4-3. The crud releasefraction for row 7 in ER Table 7.4-3 (page 7.4-9) should be 2. OOE-3 (rather than2.OOE-2).

This corrects an error in the ER. (See original reference NUREG/CR-6672 Table7.31).

The reactors proposed do not meet all requirements of 10 CFR 51.52(a);therefore this information is needed to determine compliance with 10 CFR51.52(b), which requires that. "...the statement shall contain a full description anddetailed analysis of the environmental effects of transportation of fuel and wastesto and from the reactor, including values for the environmental impact undernormal conditions of transport and for the environmental risk from accidents intransport."


The severity and release fractions in ER Table 7.4-3 are based on the values reportedin Appendix G of NUREG-1815. Table 7.31 of NUREG/CR-6672, Reexamination ofSpent Fuel Shipment Risk Estimates (ER Reference 7.4-2), specifies the crud releasefraction as 2.OOE-03 for Severity Category 7 compared to the value of 2.OOE-02 in ERTable 7.4-3. The value in Table 7.31 of NUREG/CR-6672 is the original reference andcorrect value. The value reported in the ER and used in RADTRAN is 2.OOE-02.

The value for Severity Category 7 in ER Table 7.4-3 will be changed from 2.OOE-02 to2.OOE-03 in a future revision to the ER. The value used in the RADTRAN analysis isconservative, so the environmental impact due to transportation accidents isconservative. Therefore, no further analysis is necessary.

A footnote will be included in ER Table 7.4-3 in a future revision of the ER clarifying thata conservative value of the crud release fraction was used in RADTRAN for calculatingthe impact of transportation accidents.


References for the release fractions to Appendix G of NUREG-1815 on page 7.4-3 willbe changed to Table 7.31 of NUREG/CR-6672.

The value for Severity Category 7 in Table 7.4-3 will be changed from 2.OOE-02 to2.OOE-03 in a future revision.

Enclosure 1 Page 4

The following footnote will be added to ER Table 7.4-3:

"b) A conservative value for the crud release fraction of 2.OOE-02 was used tocalculate the transportation accident risks in RADTRAN. See Appendix 7A."


Enclosure 1 Page 5


In Reference 2, the NRC staff asked PSEG for information regarding Alternative Sites,as described in Section 9.3 of the Environmental Report. The specific request was:

rTR-06: Provide a detailed transportation analysis for the four alternative sitesidentified in the ER: Site 4-1, Site 7-1, Site 7-2, and Site 7-3.

Transportation impacts for the alternative sites are not provided in ER Sections3.8, 5.7.2, 7.4, or 9.3.

10 CFR 51.50(b)(1) states that "The environmental report must include anevaluation of alternative sites to determine whether there is any obviouslysuperior alternative to the site proposed."


The routes used for the transportation of unirradiated fuel, irradiated fuel, and radwasteto and from Alternative Sites 7-1, 7-2, and 7-3 will be essentially the same as for thePSEG Site, given that these three alternate sites are located within approximately 20miles of the PSEG Site (see ER Figure 9.3-13). The local mileage traveled to reachHighway S49 from the PSEG Site, as determined by the TRAGIS code, is 12.3 miles.This distance is less than 1 percent of the total mileage for the transportation fromRichland, WA (unirradiated fuel) to Yucca Mountain, NV (irradiated fuel), and less than2 percent of the total mileage to Barnwell, SC (radwaste). Therefore, the difference inenvironmental impact from the transportation of unirradiated fuel, irradiated fuel, andradwaste between the PSEG Site and Alternative Sites 7-1, 7-2, and 7-3 is notsignificant.

Alternative Site 4-1 is located north of the PSEG Site. The route for irradiated fuel andunirradiated fuel for the PSEG Site involves traveling north to Interstate-80, which isnorth of the Alternative Site 4-1. As a result, the total mileage for transport of irradiatedfuel and unirradiated fuel for the Alternative Site 4-1 is less than the total mileage for thePSEG Site. Alternative Site 4-1 is located farther from Barnwell, SC than the PSEGSite. It is expected that the decrease in environmental impact due to irradiated andunirradiated fuel shipments will offset any increase in environmental impact fromradwaste shipments. Therefore, the difference in the environmental impact from theshipment of unirradiated fuel, irradiated fuel, and radwaste between the PSEG Site andAlternative Site 4-1 is not significant.

An evaluation of the environmental impact due to transportation of unirradiated fuel,irradiated fuel, and radwaste at Alternative Sites 7-1, 7-2, 7-3, and 4-1 indicates that thealternative sites are not obviously superior to the PSEG Site.

Enclosure 1 Page 6


Environmental Report (ER) Subsections 3.8, 5.7.2, and 7.4 will be updated to addressthe environmental impact of transportation of unirradiated fuel, irradiated fuel, andradwaste at Alternative Sites 7-1, 7-2, 7-3, and 4-1. The ER update, which will alsostate that the alternate sites are not obviously superior to the PSEG Site, will beprovided in a future revision of the ER.


Enclosure 1 Page 7


In Reference 2, the NRC staff asked PSEG for information regarding Number of TruckShipments, as described in Subsection 5.7.2.1.11 of the Environmental Report. Thespecific request was:

rTR-08: Provide the number of shipments of unirradiated fuel, irradiated fuel, andradioactive waste nomialized to the 1100 MW(e) reference reactor with acapacity factor of 0. 8. Also, state and justify the assumptions used in thecalculations when determining the aforementioned number of shipments.

The impacts in Table S-4 of 10 CFR 51.52(c) are based on an 1100 MW(e)reactor with a capacity factor of 0. 8, and specific container capacities.Normalization of the shipments is needed to compare impacts on an equal basis.


When normalizing the shipments, consider the following:1. Provide the number of unirradiated fuel shipments accounting for the initial

core loading.2. Provide the number of irradiated fuel shipments using a shipping container

capacity of 0. 5 MTU/container.3. Provide the number of radioactive waste shipments using a capacity of 82.6

ft3/shipment (2.34 m 3/shipment).


The number of shipments for each of the reactor technologies considered in the PSEGEnvironmental Report (ER) must be normalized to compare the environmental impactson an equal basis to 10 CFR 51.52 Table S-4, which is based on a reference reactoranalyzed in WASH-1238. The annual shipments are normalized based on the netelectric generation of the reference reactor relative to each reactor technologyconsidered. The electric generation and capacity factors for the reference reactor andeach reactor technology are shown in Table ESP EIS 6.0-5-1 below.

Enclosure 1 Page 8

Table ESP EIS 6.0-5-1: Net Electric Generation Normalization

Parameter 10 CFR 51.52 ABWR AP1000 U. S. EPR US-APWRTable S-4 4300 (Dual Unit)Electric Generation (MWe) 1100 1500 2300 1600 1600

Capacity Factor 0.8 0.963 0.963 0.963 0.963Net Electric Generation (MWe) 880 1444.5 2214.9 1540.8 1540.8

Normalization 1.00 0.61 0.40 0.57 0.57

The unirradiated fuel shipments are adjusted to account for the initial core loading.Each technology is assumed to have a 24-month cycle for conservatism. Therefore, itwould take the equivalent of 6 years of annual shipments to equal the number ofshipments for a full core reload. The total shipments accounting for the initial coreloading is averaged over an assumed value of 40-years for the lifetime of the plant andadded to the annual total for each reactor technology. Note that the number ofunirradiated fuel shipments for the US-APWR has been increased to be consistent withRevision 3 of the Comanche Peak COLA.

The value of the container capacity of the spent fuel shipments is scaled to 0.5MTU/container to determine the appropriate number of spent fuel shipments forcomparison to the reference LWR. The container capacity of the radwaste shipments isscaled to 2.34 m 3/shipment for comparison to the reference LWR.

The total number of annual shipments for unirradiated fuel, irradiated fuel, and radwasteare adjusted using the criteria stated in previous paragraphs for an equal comparison tothe reference reactor. The normalized number of shipments based on the net electricgeneration, accounting for the initial core loading, and different container capacities forspent fuel and radwaste are shown in Table ESP EIS 6.0-5-2.

Table ESP EIS 6.0-5-2: Normalized Number of Shipments

Reference Normalized Normalized Normalized NormalizedShipments/year LWR ABWR AP1000 U.S. EPR US-APWR

New Fuel 6 4.3 1.8 4.9 3.5Spent Fuel 60 54.5 19.5 42.7 39.8Radwaste 46 15.8 3.4 7.5 13.5

Total 112 74.6 24.7 55.1 54.8

The normalized number of annual shipments are used to determine the environmentalimpacts due to transportation of radioactive material and transportation accidents for anequal comparison to Table S-4 of 10 CFR 51.52. These impacts are the incident-freeand accident consequences reported in Environmental Report (ER) Sections 5.7 and7.4, respectively.

Enclosure 1 Page 9


A new ER subsection (Subsection 5.7.2.3) will be added to address the adjustmentspertaining to the number of shipments and the comparison of incident-freeconsequences to 10 CFR 51.52 Table S-4.

Table 5.7-13 (duplicate of Table ESP EIS 6.0-5-2) will be added to indicate thenormalized number of shipments of unirradiated fuel, irradiated fuel, and radwaste thatare used for comparison to the reference reactor.

Table 5.7-14 will be added that summarizes the incident-free radiological consequencesbased on the normalized number of shipments and compares them to the valuesindicated in 10 CFR 51.52 Table S-4 that are based on the analyzed reference reactorin WASH-1238.

Tables 5.7-5, 5.7-8, and 5.7-9 will be updated to reflect a change in the US-APWRnumber of shipments for new fuel.

The conclusion of Subsection 5.7.2 will be moved from Subsection "5.7.2.3" toSubsection "5.7.2.4". The conclusion will also be revised based on the results in Table5.7-14 rather than Table 5.7-12.

A new ER subsection (Subsection 7.4.2.4) will be added to provide environmentalimpacts from transportation accidents based on the normalized number of shipmentsand provide the comparison of accident consequences to 10 CFR 51.52 Table S-4.

Table 7.4-9 and Table 7.4-10 will be added to indicate the non-radiological accidentconsequences based on the normalized number of shipments.

Table 7.4-11 will be added to compare the bounding values from Table 7.4-9 and Table7.4-10 to the values indicated in 10 CFR 51.52 Table S-4 that are based on theanalyzed reference reactor in WASH-1238.

The conclusions made in Subsection 7.4.3 will also be revised based on the results inTable 7.4-11 rather than Table 7.4-8.


Enclosure 1 Page 10


In Reference 2, the NRC staff asked PSEG for information regarding Transportation, asdescribed in Subsection 5.7.1.8 of the Environmental Report. The specific request was:

rTR-09a: Provide a consistent value for the Plant Parameter Envelope (PPE)value for the capacity factor.

The PSEG Site Safety Analysis Report (SSAR) (page 1.3-16) states that the PPEvalue for the capacity factor is 0.963; however, ER Table 5.7-1 (page 5.7-17)states that the PPE value for the capacity factor is 0. 95.



To be consistent with the Plant Parameter Envelope (PPE), the capacity factor ischanged from 0.95 to 0.963 in the ER. This change in the capacity factor results inchanges to New Plant Bounding Values determined from scaling factors in Table 5.7-1and Table 5.7-3 of the ER. As a result, the New Plant Bounding Values reported inTables 5.7-1 and 5.7-3 are adjusted accordingly and changed in appropriate sections inSubsection 5.7.1 of the ER. These updates will be included in a future revision of theER.


Subsection 5.7.1 will be updated to reflect New Plant Bounding Values given in Table5.7-1 and Table 5.7-3 that are determined using the updated scaling factor.

Table 5.7-1 and Table 5.7-3 of the ER are revised to reflect the PPE value for thecapacity factor and the appropriate scaling from the reference reactor.


Enclosure 1 Page 11


In Reference 2, the NRC staff asked PSEG for information regarding Transportation ofRadioactive Material, as described in Subsection 5.7.2 of the Environmental Report.The specific request was:

rTR- 11: Provide capacity factors for the ABWR, APIO00, US EPR, and US-APWR.

The NRC staff needs this information to determine that the PPE value for thecapacity factor used in the transportation analysis is reasonable.

10 CFR 51.50(b)(2) states that "The environmental report may address one ormore of the environmental effects of construction and operation of a reactor, orreactors, which have design characteristics that fall within the site characteristicsand design parameters for the early site permit application, provided however,that the environmental report must address all environmental effects ofconstruction and operation necessary to determine whether there is anyobviously superior alternative to the site proposed..."



The capacity factors for the ABWR and AP1000 in Table G-1 of NUREG-1815 is 95%.The capacity factor in Table 6-13 of NUREG-1936 for the US EPR is 95%. Thecapacity factor in Table 6-3 of NUREG-1943 for the US-APWR is 93%. The PlantParameter Envelope station capacity factor for the PSEG Site is 96.3% (SSAR Table1.3-1, Item 17.6). This bounding value is used for transportation analysis of all of thereactor technologies under consideration.


Associated revisions to the ER regarding the capacity factor of 96.3% are addressed inRAI No. Env-12, Question No. ESP EIS 6.0-6.

Enclosure 1 Page 12


In Reference 2, the NRC staff asked PSEG for information regarding Fuel Enrichment,as described in Subsection 5.7.2.1.3 of the Environmental Report. The specific requestwas:

rTR-12: Provide revised text in ER Section 5.7.2.1.3, Fuel Enrichment, to reflectNRC's position that NUREG-1437 can only be used to justify enrichments greaterthan 4 weight percent for license renewals and cannot be used to justifyenrichments greater than 4 weight percent for new reactors.

The reactors proposed do not meet 10 CFR 51.52(a) (2); therefore, thisinformation is needed to determine compliance with 10 CFR 51.52(b) requiresthat: "...the statement shall contain a full description and detailed analysis of theenvironmental effects of transportation of fuel and wastes to and from thereactor, including values for the environmental impact under normal conditions oftransport and for the environmental risk from accidents in transport."

NUREG-1437 (NRC 1996; NRC 1999) cannot be used as the initial licensingbasis for new reactors. See also ESRP 5.7.2, Revision 1, page 5.7.2-3.

"...the NRC has generically considered the environmental impacts of spentnuclear fuel with U-235 enrichment levels up to 5% and irradiation levels up to62, 000 megawatt-days per metric ton and found that the environmental impactsof spent nuclear fuel transport are bounded by the impacts listed in Table S-4provided that more than 5 year has elapsed between removal of the fuel from thereactor and shipment of the fuel offsite (NRC 1996; NRC 1999). However, theseanalyses cannot serve as the initial licensing basis for new reactors."(emphasis added)

U.S. Nuclear Regulatory Commission (NRC). 1996. Generic EnvironmentalImpact Statement for License Renewal of Nuclear Plants. NUREG-1437,Washington, D.C.

U.S. Nuclear Regulatory Commission (NRC). 1999. Generic EnvironmentalImpact Statement for License Renewal of Nuclear Plants, Main Report, Section6.3 - Transportation, Table 9.1 Summary of findings on NEPA issues for licenserenewal of nuclear power plants. NUREG- 1437 Vol. 1, Addendum 1,Washington, D.C.


The use of NUREG-1437 in Environmental Report (ER) Subsection 5.7.2.1.3 asjustification for enrichment greater than 4 weight percent of a new nuclear reactor will beremoved. Further analysis of enrichment greater than 4 weight percent with the use ofRADTRAN is provided in the ER.

Enclosure 1 Page 13


The statement in ER Subsection 5.7.2.1.3 identifying NUREG-1437 as justification forenrichment greater than 4 weight percent is removed. A statement is added indicatingthat further analysis is provided in accordance with 10 CFR 51.52(b).


Enclosure 1 Page 14


In Reference 2, the NRC staff asked PSEG for information regarding Average FuelBurnup, as described in Subsection 5.7.2.1.5 of the Environmental Report. The specificrequest was:

rTR-13: Provide revised text in ER Section 5.7.2.1.5, Average Fuel Burnup, toreflect NRC's position that NUREG-1437 can only be used to justify burnupsgreater than 33, 000 MWd/MTU for license renewals and cannot be used to justifyburnups greater than 33,000 MWd/MTU for new reactors.

Full Text (Supporting Information) is identical to that for rTR-13 (immediatelyabove).


The use of NUREG-1437 in Environmental Report (ER) Subsection 5.7.2.1.5 asjustification for burnups greater than 33,000 MWd/MTU of a new nuclear reactor will beremoved. Further analysis of a burnup greater than 33,000 MWd/MTU with the use ofRADTRAN is provided in the ER.


The statement in ER Subsection 5.7.2.1.5 identifying NUREG-1437 as justification for aburnup greater than 33,000 MWd/MTU is removed. A statement is added indicatingthat further analysis is provided in accordance with 10 CFR 51.52(b).


Enclosure 1 Page 15


In Reference 2, the NRC staff asked PSEG for information regarding TransportationAccidents, as described in Subsection 7.4 of the Environmental Report. The specificrequest was:

rTR-14: Provide radionuclide inventories (CWiMTU) for the ABWR, APIO00, andUS EPR.

The NRC staff needs additional information to determine that the PPE value forthe radionuclide inventory, which is based on the US-APWR, used in thetransportation analysis in ER Table 7.4-2 (page 7.4-8) is reasonable.

10 CFR 51.50(b)(2) "The environmental report may address one or more of theenvironmental effects of construction and operation of a reactor, or reactors,which have design characteristics that fall within the site characteristics anddesign parameters for the early site permit application, provided however, thatthe environmental report must address all environmental effects of constructionand operation necessary to determine whether there is any obviously superioralternative to the site proposed..."

The reactors proposed do not meet all requirements of 10 CFR 51.52(a);therefore, 10 CFR 51.52(b) requires that: "...the statement shall contain a fulldescription and detailed analysis of the environmental effects of transportation offuel and wastes to and from the reactor, including values for the environmentalimpact under normal conditions of transport and for the environmental risk fromaccidents in transport."


The radionuclide inventory (Ci/MTU) for each of the reactor technologies at 5 yearsdecay time is shown in Table ESP EIS 6.0-10-1. The spent fuel inventory of the US-APWR is taken from the Comanche Peak Unit 3 & 4 COLA Environmental Report (ER)Table 7.4-1. The spent fuel inventory of the U.S. EPR is taken from the Calvert CliffsUnit 3 COLA ER Table 7.4-3. The ABWR spent fuel inventory is taken from the SouthTexas Project Unit 3 & 4 COLA ER Table 7.4-1. The spent fuel inventory of the AP1 000is taken from the V.C. Summer Unit 2 & 3 COLA ER Table 7.4-1. The US-APWRradionuclide inventory is the bounding design based on the value for the total Ci/MTU.

Enclosure 1 Page 16

Table ESP EIS 6.0-10-1: Comparison of 5-Year Decayed Fission ProductInventories

Nuclide(a) US-APWR U.S. EPR ABWR AP1000(CiIMTU) (Ci/MTU) (Ci/MTU) (Ci/MTU)

Am-241 1.81 E+03 1.25E+03 1.44E+03 7.27E+02Am-242m 2.04E+01 2.38E+01 3.32E+01 1.31 E+01Am-242 2.04E+01Am-243 7.45E+01 3.22E+01 5.95E+01 3.34E+01Ce-144 1.39E+04 1.52E+04 1.32E+04 8.87E+03

Pr-144 (D)Pr-144m (D)

Cm-242 6.08E+01 4.35E+01 6.22E+01 2.83E+01Cm-243 5.76E+01 3.19E+01 6.17E+01 3.07E+01Cm-244 1.25E+04 4.84E+03 1.35E+04 7.75E+03Cm-245 6.19E-01 2.25E+00 1.21 E+00Co-60 8.58E+01 7.59E+01 3.63E+03

Cs-134 6.41E+04 5.84E+04 7.76E+04 4.80E+04Cs-137 1.76E+05 1.42E+05 1.58E+05 9.31E+04

Ba-137m (D)Eu-154 1.03E+04 1.16E+04 1.56E+04 9.13E+03Eu-1 55 2.74E+03 5.73E+03 8.27E+03 4.62E+031-129 4.65E-02Kr-85 1.09E+04 1.05E+04

Pm-147 5.17E+04 3.54E+04 3.13E+04 1.76E+04Pu-238 9.51E+03 6.95E+03 1.09E+04 6.07E+03Pu-239 4.08E+02 4.24E+02 4.27E+02 2.55E+02Np-239 7.45E+01Pu-240 6.97E+02 7.24E+02 8.52E+02 5.43E+02Pu-241 1.68E+05 1.17E+05 1.35E+05 6.96E+04Pu-242 2.28E+00 3.19E+00 1.82E+00Ru-106 2.46E+04 2.05E+04 2.29E+04 1.55E+04

Rh-106 (D)Sb-125 3.39E+03 5.35E+03 7.17E+03 3.83E+03Sr-90 1.20E+05 1.03E+05 1.06E+05 6.19E+04Y-90 1.20E+05 1.03E+05 1.06E+05 6.19E+04H-3 6.49E+02

Tc-99 2.33E+01Ag-1 1Om 5.43E+01Cd-113m 4.98E+01Te-125m 8.30E+02Sm-151 6.47E+02

Total 7.93E+05 6.42E+05 7.12E+05 4.10E+05a) The nuclides labeled with a (D) are daughter products and are included with the parent

in the RADTRAN program

Enclosure 1 Page 17


None.

Enclosure 1 Page 18

PSEG Letter ND-2012-0071, dated October 19, 2012

ENCLOSURE 2

Proposed Revisions

Part 3 - Environmental Report (ER)Section 3.8 - Transportation of Radioactive Materials

Section 5.7 - Uranium Fuel Cycle and Transportation ImpactsSection 7.4 - Transportation Accidents

Marked-up Pages3.8-1

5.7-1 through 5.7-45.7-65.7-7

5.7-9 through 5.7-145.7-175.7-185.7-215.7-225.7-235.7-265.7-277.4-17.4-27.4-37.4-57.4-67.4-9

7.4-11 through 7.4-14

PSEG Site Add "and aftemative sites" perESP ApplicationPart 3, Environmental Rep• RAI No. Env-12 Question 6.04.

3.8 TRANSPORTATION OF RADIOACTIVE MATE ""/

This section describes the transportation of unirr ' ed fuel, irradiated fuel (spent nudear fuel),and radioactive waste required at the PSEG Site.

Subsection 5.7.2 also addresses 10 CFR 51.52, Environmental Effects of Transportation of Fueland Waste, (a)(1) through (a)(5) regarding use of Table S-4 to characterize both the impacts ofradioactive materials transportation and provide an analysis of the radiological impacts fromincident-free transportation of these materials. Section 7.4 addresses postulated radiologicaltransportation accidents.

3.8.1 TRANSPORTATION OF UNIRRADIATED FUEL

New fuel assemblies are transported to the PSEG Site from a fuel fabrication facility inaccordance with Department of Transportation (DOT) (49 CFR Parts 173, Shippers - GeneralRequirements for Shipments and Packagings, 178, Specifications for Packagings, and 397,Transportation of Hazardous Materials; Driving and Parking Rules) and NRC regulations (10CFR Part 71, Packaging and Transportation of Radioactive Malteal). The fuel assemblies areshipped by truck to the PSEG Site shortly before they are required. The container designs,shipping procedures, and transportation routing will be in accordance with DOT and NRCregulations and depend on the requirements of the suppliers providing the fuel fabricationservices. The truck shipments will not exceed 73,000 pounds as governed by federal or stategross vehide %eight restrictions.

3.8.2 TRANSPORTATION OF IRRADIATED FUEL

Spent fuel assemblies are discharged from each unit and will remain in spent fuel poolsassociated with the new units at least five years while short half-life isotopes decay. The newplant has sufficient spent fuel capacity to ensure that irradiated fuel can be stored for at leastfive years before being removed from the spent fuel pool. After a sufficient decay period, thefuel will be removed from the pool and packaged in spent fuel shippinglstorage casks, licensedin accordance with 10 CFR 72, Ucensing Requirements for the Independent Storage of SpentNuclear Fuel and High-Level Radioactive Waste, and Reactor-Related Greater Than Class CWaste, and transferred either to an independent spent fuel storage installation facility on-site oran off-site disposal facility. Packaging of the fuel for off-site shipment will comply with applicableDOT (49 CFR 173, 49 CFR 178, and 49 CFR 397) and NRC regulations (10 CFR 71) fortransportation of radioactive material. By law, the Department of Energy (DOE) is responsiblefor spent fuel transportation from reactor sites to a repository (Nuclear Waste Policy Act of 1982,as amended). DOE will determine the mode of transport

3.8.3 TRANSPORTATION OF RADIOACTIVE WASTE

As described in Subsection 3.5.3, low-level radioactive waste is packaged to meettransportation and disposal site acceptance requirements. Waste packaging for off-siteshipment will comply with applicable DOT (49 CFR 173 and 49 CFR 178) and NRC regulations(10 CFR 71) for transportation of radioactive material. The packaged waste will be stored on-site on an interim basis before being shipped off-site to a licensed processing, storage, ordisposal facility. Radioactive waste is shipped off-site by truck.

Rev. 13.8-1

Enclosure 2 Page 1

PSEG SiteESP Application

Part 3, Environmental Report

5.7 URANIUM FUEL CYCLE AND TRANSPORTATION IMPACTS

Subsection 5.7.1 addresses the environmental impacts from the uranium fuel cycle.Subsection 5.7-2-1 addresses the conditions in subparagraphs 10 CFR 51.52(a)(1) through (5)regarding use of Table S-4 to characterize the impacts of radioactive materials transportation inthis Environmental Report. An analysis of the transportation effects was performed because notall the conditions set forth in 10 CFR 51.52(a) are met. Subsection 5.7.2.2 addresses theincident-free transportation of radioactive materials. Postulated accidents due to transportationof radioactive materials are discussed in Section 7.4.

5.7.1 URANIUM FUEL CYCLE IMPACT

This subsection discusses the environmental impacts from the uranium fuel cycle for the newplant at the PSEG Site. The evaluations of potential environmental effects of the new plant arebased on bounding information from the PPE. The reactor types considered are the AdvancedBoiling Water Reactor (ABWR), AP 1000, U.S. Evolutionary Power Reactor (U.S. EPR), andU.S. Advanced Pressurized Water Reactor (US-APWR). A description of the development andintended use of the PPE is provided in Section 1.3 of the SSAR.

The uranium fuel cycle is defined as the total of those operations and processes associated withprovision, utilization, and ultimate disposition of fuel for nuclear power reactors.

The regulations in 10 CFR 51.51, Uranium Fuel Cycle Environmental Data - Table S-3,paragraph (a), state:

'Every environmental report prepared for the construction permit stage or early sitepermit stage or combined license stage of a light-water-cooled nuclear power reactor,and submitted on or after September 4, 1979, shall take Table S-3, Table of UraniumFuel Cycle Environmental Data, as the basis for evaluating the contribution of theenvironmental effects of uranium mining and milling, the production of uraniumhexafluoride, isotopic enrichment, fuel fabrication, reprocessing of irradiated fuel,transportation of radioactive materials and management of low-level wastes and high-level wastes related to uranium fuel cycle activities to the environmental costs oflicensing the nuclear power reactor. Table S-3 shall be included in the environmentalreport and may be supplemented by a discussion of the environmental significance ofthe data set forth in the table as weighed in the analysis for the proposed facility."

NRC Table S-3 is used to assess environmental impacts associated with the uranium fuel cycle.Its values are normalized for a reference 1000 megawatts electric (MWe) light water reactor(LWR) at an 80 percent capacity factor. The 10 CFR 51.51(a), Table S-3 values are reproducedas the Reference LWR column in Table 5.7-1. For the new plant, the bounding technology withrespect to power level is a dual unit AP1000. A scale factor based on MWe and capaciis used to develop S-3 values for a dual unit AP1000. For this analysis the scale factor i 2-73.comparison of the bounding technology to the reference LWR is induded in Table 5.7-1.

Specific categories of natural resource use are included in NRC Table S-3 (and duplica td in

Table 5.7-1). These categories relate to land use, water consumption and thermal effluents,radioactive releases, burial of transuranic and high-level and low-level wastes, and ra iation

doses from transportation and occupational exposures. Ii n deveo in N ab-nd the NRC5T7a1bleD~eplACe wih-.)77. T / Rev. 1

57-1 per RAI No. Env-12Question 6.0-6.

Enclosure 2 Page 2



initially considered two fuel cycle options, which differed in the treatment of spent fuel removedfrom a reactor. No recycle treats all spent fuel as waste to be stored at a federal wasterepository. Uranium only recycle involves reprocessing spent fuel to recover unused uraniumand return it to the system for reuse. Neither cycle involves the recovery of plutonium. Thecontributions in NRC Table S-3 resulting from reprocessing, waste management, andtransportation of wastes are maximized for both of the two fuel cycles (uranium only recycle andno recycle); that is, the cycle that results in the greater impact is used.

The United States does not currently reprocess spent fuel, therefore only the no recycle optionis considered. As described in NUREG-1555, natural uranium is mined in either open-pit orunderground mines, or by an in situ mining process. In situ leach mining, the primary form ofmining in the United States today, involves injecting a lixiviant (leaching solution) into theuranium ore body to dissolve uranium and then pumping the solution to the surface for furtherprocessing. The ore or in situ leach solution is transferred to mills where it is processed toproduce yellow-cake (UMOB). A conversion facility prepares the uranium oxide by converting it touranium hexafluoride (UF5), which is then processed by an enrichment facility to increase thepercentage of the more fissile isotope uranium-235 (U-235) and decrease the percentage of thenonfissile isotope uranium-238. At a fuel-fabrication facility, the enriched uranium is converted touranyl acetate (U0 2 ). The U02 is pelletized, sintered, and inserted into tubes to form fuelassemblies. The fuel assemblies are placed in the reactor to produce power. When the contentof the uranium-235 reaches a point where the nuclear reactor has become inefficient withrespect to neutron economy, the fuel assemblies are withdrawn from the reactor. After on-sitestorage for sufficient time to allow for short-lived fission product decay and to reduce the heatgeneration rate, the fuel assemblies will be transferred to a federal repository for internment.Disposal of spent fuel elements in a repository constitutes the final step in the no-recycle option.

The following assessment of the environmental impacts of the fuel cycle for the boundingtechnology at the PSEG Site is based on the values in NRC Table S-3 and the NRC's analysisof the radiological impacts from radon-222 and technetium-99 provided in NUREG-1437.NUREG-1437 provides a detailed analysis of the environmental impacts from the uranium fuelcycle. Although NUREG-1437 is specific to impacts related to license renewal, the information isrelevant to this review because the reactor technologies being considered use the same type offuel.

The fuel impacts in NRC Table S-3 are based on a reference 1000-MWe LWR operating at anannual capacity factor of 80 percent for a net electric output of 800 MWe. When evaluating thenew plant, NUREG-1555, Section 5.7.1, directs that the impacts in Table S-3 are scaled to thenet electric output for the new plant. For the dual unit AP1 000, the surrogate AP1000 from

eG-1815. Appendix G is used. This has a gross electric output of 1150 MWe and aH6 rcent capac• ctor for a single unit The combined gross elec•tie t for tw Replace wh "2216'

0 MWe and 96 ercent capacity factor for a net electric output-3 , .2pe ,wRnI/N. Ev-.pr - per RAI- No. -nv-12esuits are pro*"e in Table 5.7-1. f t Question 6.0-6.

shown in able 5.7-2, the new plant may require more than 35 metric tons of uranium' T .er yr. This ble also shows the fuel cycle uranium requirements. The requirements for naturalFr, enric ed UF6, UsQa and separative work units (SWU) are based on the approach used ine ldah National Engineering and Environmental Laboratory, Early Site Permit Environmentalport ections and Supporting Documentafion, (Reference 5.7-2), i.e., scaled from the

r ee ce plant using the annual fuel load. The uranium requirements may exceed 35 MTU

Replace with "96.3" per RAI Rev. 1-. 5.7-2

No. Env-12 Que on 6.0-6.

Enclosure 2 Page 3



because the generating capacity is greater than the reactor designs that were considered whenNUREG-1437 was issued. At least two of the reference 1000 MWe LWRs may be required toprovide the generating capacity of the new plant (Reference 5.7-2).

Recent changes in the fuel cycle may reduce environmental impacts in the United States. TheNRC calculated the values in NRC Table S-3 from industry averages for the performance ofeach type of facility or operation associated with the fuel cycle. The NRC chose assumptions sothat the calculated values are not underestimated. This approach was intended to ensure thatthe actual values are less than the quantities shown in NRC Table S-3 for all LWR nuclearpower plants within the widest range of operating conditions. Since NRC Table S-3 waspromulgated, changes in the fuel cycle and reactor operations have occurred. For example, theestimated quantity of fuel required for a year's operation of a nuclear power plant can nowreasonably be calculated assuming a 60-yr lifetime (40 yr of initial operation plus a 20-yr licenserenewal term). This is described in NUREG-1 437, for both BWRs and PWRs, and the highestannual requirement, 35 MTU made into fuel for a BWR, was used as the basis for the referencereactor year.

Since the original estimates in 1979 were made for Table S-3, a number of fuel managementimprovements have been adopted by nuclear power plants to achieve higher performance andto reduce fuel and separative work (enrichment) requirements. These improvements areestimated to reduce the annual fuel requirement by 10 to 15 percent.

In addition, the Table S-3 estimates for enrichment are based on the gaseous diffusion process,which has been used in the United States since the earliest days of the nuclear power program.The largest impacts of the gaseous diffusion process are attributable to the large requirementfor electric energy to run the plant (especially to the assumption that the electricity will comefrom coal-fired power plants) and to the large amount of cooling water used in the gaseousdiffusion process equipment. The centrifuge process uses 90 percent less electrical energy andtherefore, has far lower impacts attributable to coal-fired power plants and the use of coolingwater. The assumption of continued use of United States diffusion enrichment services ensuresthat environmental impacts are not underestimated.

Factoring in changes to the fuel cycle suggests that the environmental impacts of mining and tailmillings could drop to levels below those in NRC Table S-3. Section 6.2 of NUREG-1437discusses the sensitivity of these changes in the fuel cycle on the environmental impacts.

5.7.1.1 Land Use "Replacewith "127 ha (313 ac.)' perRA:INo. rEnv-12 Question 6.0-6.

As shown in Subsection 6.2.2 of NUREG-1437, the total annual land requirements for the fuelcycle supporting the 1000-MWe LWR are 46 hecta es (ha) (113 ac.). This includes 5 ha (13 ac.)that are permanently committed, and 41 ha (100 a .) that are temporarily committed. The totalannual land requirements for the fuel cycle support g the new plant are scaled up from thereference reactor and provided in Table 5.7-1. Th - M 7 ent (both temporary andpermanent land commitments) for the new plant i 4e-( a9-e •A temporary landcommitment is a commitment for the life of the sp ij 1• nt (e.g., a mill, enrichmentplant, or succeeding plants). Following decommissioning, the land could be released forunrestricted use. Permanent commitments represent land that may not be released for use afterdecommissioning because the decommissioning does not result in the removal of sufficient

Rev. 15.7-3

Enclosure 2 Page 4

Replace with "224 ha (554A ac.)" per Rpaevih"7RAI No. Env-12 Question 6.0-6. PSEG Site per RAI No. Env-12

ESP Application Question 6.0-6.Part 3, Environmental Repo

radioactive material to meet the mits of 10 CFR 20, Subpart for release of an area forunrestricted use.

In comparison to the new plant I nd use values in Table 5.7, a coal plant of 1000 MWe(800 MWe net) capacity using st mined coal require 00 ac.) per year for(NUREG-1555, Subsection - factor f-23-3 sows thata2300MW (24.9&MWe net) coal plant requir- a rest1plant are SMALL.Relcwih215

5-7.1.2 Water Use p RAIsto n v-0 2

According to Section 6.2.2.7 of NUREG-1437, principal water use for the fuel cycle is thatrequired to remove waste heat from the power stations supplying electricity to the enrichmentprocess. NUREG-1437 indicates that on a thermal-effluent basis, annual discharges from thenuclear fuel cycle are 4 percent of those from the reference 1000-MWe LWR using once-through cooling. The consumptive water use of 0.6 x 10f m3/yr (1.6 x 10' gallonslyr) is 2 percentof that from the model 1000-MWe LWR using cooling towers. The maximum consumptive wateruse (assuming that all plants supplying electrical energy to the nuclear fuel cycle used coolingtowers) are 6 percent of that of the model 1000-MWe LWR using cooling towers. In NUREG-1437, it was determined that these combinations of thermal effluents are SMALL. The fuel cyclewater use for the new plant is provided in Table 5.7-1.

5.7.1.3 Fossil Fuel Impacts

Electric energy and process heat are required during various phases of the fuel cycle process.The electric energy is usually produced by the combustion of fossil fuel at conventional powerplants. Electric energy associated with the fuel cycle represents 5 percent of the annual electricpower production of the reference 1000 MWe LWR.

Process heat is primarily generated by the combustion of natural gas. As concluded in NUREG-1437, this gas consumption, if used to generate electricity, is less than 0.4 percent of theelectrical output from the reference reactor. As a result, the direct and indirect consumption ofelectrical energy for fuel cycle operations is SMALL relative to the power production of the newplant.

The natural gas consumption associated with the fuel cycle for the new plant is greater than thereference reactor because the new plant has a significantly higher generating capacity.However, if a comparative basis is established by scaling the reference reactor to the new plant,this figure remains less than 0.4 percent of the new plant output

5.7.1.4 Chemical Effluents

The quantifies of liquid, gaseous and particulate discharges associated with the fuel cycleprocesses are given in NRC Table S-3 (Table 5.7-1) for the reference 1000 MWe LWR. Thequantities of effluents for the new plant are approximately three times those in NRC Table S-3(Table 5.7-1). The principal effluents are SOx, NOx, and particulates. Based on the 1997 AnnualReport of the Council on Environmental Quality, Chapter 5 Air Quality, the new plant emissionsconstitute a small fraction of the national sulfur and nitrogen oxide annual emissions (Reference5.7-1).

Rev. 15.7-4

Enclosure 2 Page 5



risks from technetium-99 are from exposure of the gastrointestinal tract and kidney. There is asmall risk from exposure to the whole body. Applying the organ-specific dose weighting factorsfrom 10 CFR Part 20 to the gastrointestinal tract and kidney doses, the total-body 100-yr dosecommitment from technetium-99 was estimated to be 1 person-Sv (100 person-rem) for the1 000-MWe LWR scaled model. This is scaled by the electric power rating for the new plant. Thenew plant scaled releases and doses are provided in Table 5.7-3.

As stated in Section 5.7.1 of NUREG-1555, radiation may cause cancers at high doses and highdose rates. Currently there are no data that unequivocally establish the occurrence of cancerfollowing exposure to low doses or low dose rates, below 100 millisieverts (mSv)(10,000 mrem). However, radiation protection experts conservatively assume that any amountof radiation may pose some risk of causing cancer or a severe hereditary effect and that the riskis higher for higher radiation exposures. Therefore, a linear, no-threshold dose response modelis used to describe the relationship between radiation dose and detriments such as cancerinduction. A report by the National Research Council (2006), the BEIR Vii report (Reference5.7-4), supports the linear, no-threshold dose response model. Simply stated, any increase indose, no matter how small, results in an incremental increase in health risk. This theory isaccepted by the NRC as a conservative model for estimating health risks from radiationexposure, recognizing that the model probably overestimates those risks.

The radiological impacts associated with the new plant are provided in Table 5.7-3. Based onthis model, risk to the public from the uranium fuel cycle can be estimated using the nominalprobability coefficient for total detriment (730 fatal cancers, nonfatal cancers, or severehereditary effects per 10,000 person-Sv [1.OE+06 person-rem]) from the InternationalCommission on Radiation Protection (IC cation 60 and the estimated Total EffectiveDose Equivalent (TEDE) from Table 5.7- (- _on-rem/yr). These values yieldapproximately 1.7 fatal cancers, nonfatal cancers re hereditary effects annually. Thisrisk is small compared to the number of fatal cancers, nonf~a_-ccers or severe ,Rla, with ,,,27,effects estimated to occur in the U. S. population annuay from expos naturR with "2radiation using the same risk estimation methods. :per RAI No. Env- 21

:Question 6.0-6.Based on these analyses, the environmental impacts of radioactive effluents fromare SMALL.

5.7.1.6 Radioactive Wastes

The quantities of buried radioactive waste material (low level, high level, and transuranicwastes) are specified in Table S-3. For low-level waste disposal at land burial facilities, the NRCindicates in Table S-3 that there are no significant radioactive releases to the environment. Forhigh level and transuranic wastes, the NRC states that these are buried at a federal repositoryand that no release to the environment is associated with such disposal, although it has beenassumed that all of the gaseous and volatile radionuclides contained in the spent fuel arereleased to the atmosphere before the disposal of the waste.

There is some uncertainty regarding the limits for off-site releases of radionuclides from theeventual repository site. Nevertheless, it is expected that future standards will be similar to thecurrent post-closure individual protection standard in 10 CFR 63 Subpart L, Postclosure PublicHealth and Environmental Standards. This standard requires that the maximally exposed

Rev. 15.7-6

Enclosure 2 Page 6



individual receive an annual dose of no more than 0.15 mSv (15 mrem) for 10,000 yr followingdisposal; and 1.0 mSv (100 mrem) after 10,000 yr, but within the period of geologic stability.

For the reasons stated above, the e Replace with "16.6" aste disposal are SMALL.

5.7.1.7 Occupational Dose per RAI No. Env-1-2Question 6.0-6.

As stated in Section 6.2.2.3 of NUR - , pational dose for the reference1000 MWe reactor attributable to all phases of the fuel cycle' 6 person-Sv (600 person-rem).The fuel cycle for the new plant is similar to the fuel cycle of e reference reactor and theannual occupational dose for all phases of the fuel cycle ca be determined b klizing therated power of the new plant to the reference reactor. e scale foccupational dose for the fuel cycle for the new plant i 4-, erson-Sv 4640 -rem).However, on a per MWe basis, the dose is the same. oses will be les f 10 CFR 20limit of 0.05 Svlyr (5 remfyr) to any individual worker. The environmen impact aimoccupational dose is SMALL.

occpatona doe i SMLL eplace with "1660" Replace with `237T'.... ..1..rnpotto ,:per RAI No. Env-1 2 : -•per RAJ No. Env-1 25.7.1.8 T 6.0-6. Question 6.0-6.

As indicated in NUREG 1555, the transpo lion ose or ers and the public totals0.025 person-Sv (2.5 person-rem) annually for the reference 1000 MWe LWR per Table S-3.The scaled occupational dose for the new plant is provided in Table 5.7-1. For comparison, theestimated collective dose from natural background radiation to the population within 80 km (50mi.) of the PSEG Site is 8312 person-Svlyr (831,200 person-rem/yr). This is based on a 2010population of 5,460,955, as shown in Table 2.5-7 of this ER, and an average individual dose of152 mremlyr in New Jersey (Reference 5.7-3). On the basis of this comparison, environmentalimpacts of transportation are SMALL.

5.7.1.9 Fuel Cycle

Only the no recycle option is considered here because the United States does not currentlyreprocess spent fuel. The data provided in Table S-3, however, includes the maximum recycleoption impact for each element of the fuel cycle. The analysis of the uranium fuel cycleperformed and the environmental impacts, as compared to Table S-3 impacts, are not affectedby the specific fuel cycle selected.

5.7.1.10 Conclusion

Using an evaluation process specified in NUREG-1437, PSEG evaluated the environmentalimpacts of the uranium fuel cycle, considered the effects of radon-222 and technetium-99, andappropriately scaled the data for the new plant. Based on this evaluation, the environmentalimpacts of the uranium fuel cycle are SMALL.

5.7.2 TRANSPORTATION OF RADIOACTIVE MATERIALS

This subsection addresses the environmental impacts of incident-free transportation ofradioactive materials from the PSEG Sitezor the purposes of the evaluations in this subsectionit is assumed that all shipments of fuel an cte waste are b truck.

Add "and alternative sites" perRAI No. Env-12 Question 6.0-4. Rev. 1

5.7-7

Enclosure 2 Page 7



The detailed analyses required by 10 CFR 51.52 is performed using the TRAGIS (Reference5-7-6) and RADTRAN (References 5-7-8 and 5.7-9) computer codes. The results of theseanalyses are summarized in Subsection 7.4.3. Input and output streams for these codes arecontained in Appendix 7A_Table 5.7-4 summarizes the characteristics of the reference reactor specified in 10 CFR 51.52Table S-4, along with the characteristics of the reactor technologies under consideration.

5.7.2.1.1 Reactor Core Thermal Power

Subparagraph 10 CFR 51.52(a)(1) requires that the reactor core thermal power level not exceed3800 MWt. The thermal power levels for all the reactors being considered for the PSEG Siteexceed 3800 MWt, except for a single unit AP1000. Therefore, in accordance with 10 CFR51.52(b), further analysis is provided in this subsection.

5.7.2.1.2 Fuel Form

Subparagraph 10 CFR 51.52(a)(2) requires that the reactor fuel be in the form of sintered U0 2pellets. All of the technologies being considered for the new plant use a sintered U0 2 pellet fuelform.

5.7.2.1.3 Fuel Enrichment

Subparagraph 10 CFR 51.52(a)(2) requires that the reactor fuel have a U-235 enrichment notexceeding 4 percent by weight. The maximum fuel enrichment for new plant technologiesexceeds 4 percent U-235 by weight, but is less than 5 percent by weight. Paragraph 10 CFR51.52 (b) states that, for reactors not meeting the conditions of paragraph 51.52 (a), a fulldescription and detailed analysis of the environm - s t, and

~~pacitwich to nurthtrad atcdfc can legtr o appistespet 5 rbfuel e bcing uptovcd ferom~

,,, aspent fuel pook. Therefore, the now plant technolo~gie- meetthiscond-ition

5.7.2.1.4 Fuel Encapsulation

Subparagraph 10 CFR 51.52(a)(2) requires that the rea or fuel pellets be encapsulated inzircaloy rods.

The AP1000 uses ZIRLO cladding. However, the re irement for zircaloy has been modified by10 CFR 50.46 to allow the use of ZIRLO.

The U.S. EPR uses M5 cladding, which is not co ered in 10 CFR 50.46. The NRC hasapproved the use of MS cladding by approving ramatome ANP topical report BAW-10227P-A,Evaluation of Advanced Cladding and Structur I Material (M5) in PWR Reactor Fuel (Reference5.7-5).

Replace with "in accordance with 10CFR 51.52(b), further analysis is

~Rev. 1

provided in this subsection." per RAIRNo. Env- 12 Question 6.0-8.

Enclosure 2 Page 8

Add'The maximum bumnup fornew Replace with "inplant technologies exceeds

33,OOM~/MT. pr RI N. PEG iteaccordance with 10 CFRQ33,000Md/MTU. per RAI No. ESP Application .51.52(b), further analysis isEnv-1 2 Question 6.0-9.Env-12 art 3, Environmental Report :provided in this

subsection." per RAI No.5.7.2.1.5 Average Fuel Bumup :Env-2 Question 6.0-9.

Subparagraph 1 FIR 51.52(a)(3) requires that the average bumup not exceed33,000 MWdIMT • Paragraph 10 CFR 51.52 (b) states that, for reactors not meetin econditions of paragraph 51.52 (a), a full description and detailed analysis of the envieffects of trans ortation of fuel and wastes to and from the reactor shall be rovide

l~-~o~ pro~adb teNRC up to 62,000 MWd•MTU, provided higher- bumup fuol i2 co.led•.foat Icat 5 r i b ifere bcing shipped Aff Gitc The new plant wlfl haiv sufficicnt spIIt fuel toWrage

pcit t ccu .~rc fo nwntha irreadiatcd fuel canbot 5 beforenotho 60934 fuel poo.Thrl oe. the- ",' olant technologies meet this condition.

(.

5.7.2.1.6 Time after Discharge of Irradiated Fuel before Shipment

Subparagraph 10 CFR 51.52(a)(3) requires that no irradiated fuel assembly be shipped until atleast 90 days after it is discharged from the reactor. Table S-4 assumes 150 days of decay timeprior to shipment of any irradiated fuel assemblies. For the reactor technologies beingconsidered, 5 yrs is the minimum decay time expected before shipment of irradiated fuelassemblies. The 5-yr minimum time is supported additionally by two current practices. One isper contract with DOE, which has ultimate responsibility for the spent fuel. Five yr is theminimum cooling time specified in 10 CFR 961, Standard Contract for Disposal of SpentNuclear Fuel and/or High-Level Radioactive Waste, Appendix E. The other practice is that theNRC specifies 5 yrs as the minimum cooling period when they issue certificates of compliancefor casks used for shipment of power reactor fuel. The new plant will have sufficient spent fuelstorage capacity to ensure that irradiated fuel can be stored for at least 5 yr before beingremoved from the spent fuel pool. Therefore, the new plant technologies meet this condition.

5.7.2.1.7 Transportation of Unirradiated Fuel

Subparagraph 10 CFR 51.52(a)(5) requires that unirradiated fuel be shipped to the reactor siteby truck. Unirradiated fuel shipments for the new plant will be by truck. Table S-4 includes acondition that the truck shipments not exceed 73,000 pounds (lb.) per truck as governed byfederal or state gross vehicle weight restrictions. The unirradiated fuel shipments to the PSEGSite will comply with federal, state, and local weight restrictions.

5.7.2.1.8 Radioactive Waste Form and Packaging

As specified in paragraph 10 CFR 51.52(a)(4), with the exception of spent fuel, radioactivewaste shipped from the reactor will be packaged and shipped in a solid form.

5.7.2.1.9 Transportation of Irradiated Fuel

Subparagraph 10 CFR 51.52(a)(5) allows for truck, rail, or barge transport of irradiated fuel. Forthe impact analysis described in Subsection 5.7.2, all irradiated fuel shipments are made usinglegal-weight trucks.

Rev. 15.7-10

Enclosure 2 Page 9



5-7.2.1.10 Transportation of Radioactive Waste

Subparagraph 10 CFR 51.52(a)(5) requires that the mode of transport of low-level radioactivewaste be either truck or rail. For the impact analysis described in Subsection 5.7.2, it isassumed that all radioactive waste shipments will be made using legal-weight trucks.Radioactive waste will be shipped in compliance with federal, state, and local weightrestrictions.

572.1.11 Number of Truck Shipments

As a method of limiting the environmental impact of transportation, Table S-4 limits trafficdensity to less than one truck shipment per day or three rail cars per month. The number oftruck shipments required has been estimated assuming that all radioactive materials (fuel andwaste) are received at the site or transported off-site via truck. The total number of truckshipments for the new plant is the sum of the unirradiated (new) fuel, irradiated (spent) fuel, andradwaste shipments.

A review of the unirradiated fuel shipment requirements for the technologies being consideredindicates that the bounding case is the U.S. EPR with 7.5 shipments/yr. The annual unirradiatedfuel shipment requirements are summarized in Table 5.7-5.

For the purposes of this evaluation, the annual quantity of irradiated fuel is assumed to be thesame as the annual quantity of unirradiated fuel. Shipping cask capacity assumptions are basedon current shipping cask designs. The irradiated fuel cask capacity is assumed to be 4000 lb. ofuranium (1.8 MTU) consistent with NUREG-1 811, Environmental Impact Statement for an EadySite Permit (ESP) at the North Anna ESP Site; NUREG-1 815, Environmental Impact Statementfor an Early Site Permit (ESP) at the Exelon ESP Site; and NUREG-1817, Environmental ImpactStatement for an Early Site Permit (ESP) at the Grand Gulf ESP Site. The irradiated fuelshipments are summarized in Table 5.7-6. As shown in Table 5.7-6, the bounding case is 24.8shipments/yr.

For the purposes of this evaluation, each radwaste container is assumed to be shippedseparately, that is, one container per truck. The total number of radwaste containers isdetermined by assuming that dry active waste (DAW) is shipped in Sea-Land containers with aninternal useable volume of 28.32 m3 (1000 cubic feet Ift]), and all other waste (e.g., resins,filters, etc.) are shipped in high integrity containers (HICs) with a useable intemal volume of 2.55m3 (90 ft). The annual new radwaste shipment requirements are summarized in Table 5.7-7.The annual truck shipment totals are summarized in Table 5.7-8.

7.2.2 Incident-Free Transportation Impacts Analysis

En ironment impacts of incident-free transportation of fuel are discussed in this subsection.Tra sportation accidents are discussed in Section 7.4.

5.7.2 .1 Transportation of Unirradiated Fuel

Table -4 of 10 CFR 51.52 includes conditions related to radiological doses to transport workersand me bers of the public along transport mutes. These doses, based on calculations inWASH- 38 (Reference 5.7-10), are a function of the radiation dose rate emitted from the

A5.7-11 Rev. 1

ubsec on) pr RI No1

Enclosure 2 Page 10

,Add "The difference in incident- 4PSEG Site :free consequences due to 4

ESP Application :transportation of unirradiated fuelPart 3, Environmental Report :to the Alternative Sites is not 4

.4!sigtnificant due to the smallunirradiated fuel shipments, the number of exposed individuals/sshipment, the time of transit (including travel and stop times), n differences in mileage betweenec'which the individuals are exposed. the Alternative Sites and the

assumed fabrication facility." perCalculation of worker and public doses associated with a ual sh RAI No Env-12 Question 6.0-4.were performed using the TRAGIS (Reference 5.7-6) a RADT .5.7-9) computer codes. One of the key assumptions irASH-1238 (Reference 5.7-10)for thereference LWR unirradiated fuel shipments is that th radiation dose rate at 1 m (3.3 ft) fromthe transport vehicle is 0.1 milliremlhr. This assum ion is reasonable for the new planttechnologies because the fuel materials will be I -dose rate enriched uranium and will bepackaged similarly.

For unirradiated fuel shipments, highway rou s are analyzed using the muting computer codeTRAGIS (Reference 5.7-6). It is assumed t all unirradiated fuel shipments come from the fuelfabrication facility located in Richland, Wa ington (WA), which is the furthest from the PSEGSite. The commercial route setting was ed to generate highway routes generally used bycommercial trucks. The distance from t PSEG Site to Richland, WA is 2733 mi. Thepopulation summary module of the T GIS (Reference 5.7-6) computer code is used todetermine the exposed populations thin 800 m (2625 ft.) of either side of the route.

The per trip dose values are com mned with the average annual number of shipments ofunirradiated fuel to calculate an al doses to the public and workers for comparison to Table S-4 dose values. The number of hipments per year is obtained from Table 5.7-5. The results forthe unirradiated fuel shipme based on the RADTRAN (References 5.7-8 and 5.7-9) analysesare provided in Table 5.7-9.

5.7-222 Transportation of Irradiated Fuel

The environmental impacts of transporting spent fuel from the PSEG Site to a spent fueldisposal facility assume Yucca Mountain, Nevada (NV) as a possible location for a geologicrepository. The impacts of the transportation of spent fuel to a possible repository in NVprovides a reasonable determination of the transportation impacts to a monitored retrievablestorage facility because of the distances involved and the representative exposure of membersof the public in urban, suburban, and rural areas (NUREG-1811, NUREG-1815, NUREG-1817).

Incident-free transportation refers to transportation activities in which the shipments reach theirdestination without releasing any radioactive cargo to the environment. Impacts from theseshipments are from the low levels of radiation that penetrate the heavily shielded spent fuelshipping cask. Radiation doses occur to the following:

" Persons residing along the transportation corridors between the PSEG Site and theproposed repository

" Persons in vehicles passing a spent fuel shipment* Persons at vehicle stops for refueling, rest, and vehicle inspections" Transportation crew workers

This analysis is based on shipment of spent fuel by legal-weight trucks in casks withcharacteristics similar to casks currently available (i.e., massive, heavily shielded, cylindricalmetal pressure vessels). Each shipment is assumed to consist of a single shipping cask loaded

Rev. 15.7-12

Enclosure 2 Page 11

Add 'rThe difference in incident-free

PSEG Site : consequences due to transportationESP Application of irradiated fuel from the Alternative

Part 3, Environmental Repo Sites is not significant due to thesmall differences in mileage between

on a modified trailer. These assumptions are consistent with a the Alternative Sites in comparison toenvironmental impacts of spent fuel transportation in Adden u the distance of travel to the assumeddiscussed in NUREG-1437, these assumptions are conservi .assumptions involve rail transportation or heavy-haul truck th repository. per RAI No. -Env-12spent fuel shipments. / Question 6.0-4.

The transportation route selected for a shipment determi es the total potentially exposedpopulation and the expected frequency of transportation related accidents. For trucktransportation, the route characteristics most important o the risk assessment include the totalshipping distance between each origin-destination pair sites and the population density alongthe route.

For irradiated fuel, it is assumed that all irradiated fu is sent to the site of the proposed YuccaMountain repository. The distance from the PSEG S' to the proposed repository ,asdetermined to be 2780 mi. by the TRAGIS (Referen 5.7-6) computer code for a highwayroute-controlled quantity (HRCQ).

Routing and population data used in RADTRAN ( eferences 5.7-8 and 5.7-9) for truckshipments are obtained from the TRAGIS (Refe nce 5.7-6) computer code. The populationdata in the TRAGIS (Reference 5.7-6) computer code is based on the 2000 U.S. census. Allspent fuel shipments are transported by legal-w ight trucks to the potential Yucca Mountain siteover designated HRCQ routes.

Although shipping casks have not been desig ed for the advanced LWR fuels, the advancedLWR fuel designs are not significantly differee from existing LWR designs. Current shippingcask designs are used for analysis.

The population doses are calculated by m tiplying the number of spent fuel shipments per yearby the per-shipment doses. The numbers f shipments per year are obtained from Table 5.7-6.The results for the irradiated fuel shipme based on the RADTRAN (References 5.7-8 and 5.7-9) analyses are provided in Table 5.7-10.

5.7.22.3 Transportation of Radwaste

This subsection provides the environmental impacts of transporting radwaste from the PSEGSite to the repository in Barnwell, South Carolina (SC).

Incident-free transportation refers to transportation activities in which shipments reach theirdestination without releasing any radioactive cargo to the environment. Impacts from theseshipments are from the low levels of radiation that penetrate the radwaste shipping containers.Radiation doses occur to the following:

" Persons residing along the transportation corridors between the PSEG Site and theproposed repository

" Persons in vehicles passing a radwaste shipment• Persons at vehicle stops for refueling, rest, and vehicle inspections• Transportation crew workers

Rev. 15.7-13

Enclosure 2 Page 12

-Add "The difference in

Add Insert B (new PSEG Site free consequences du

Subsection) per RAI No. ESP Application transportation of radw

Env-12 Question 6.0-5. Part 3, Environmental Report the Alternative Sites is;significant due to the s

This analysis is based on shipment of radwaste by legal-weight truck i differences in mileagecontainers or HICs similar to those currently available. Each shipme t i the Alternative Sites aa single shipping container. assumed radwaste reiThe transportation route selected for a shipment determines the t /t I per RAI No. Env-12 0

population and the expected frequency of transportation-related ccide 6.0-4...nar. ' . tics most important to the ri assessment include the total

ship Add "An evaluation of the -gin-destination pair of sites nd the population density alongthe :environmental impact due:

:to transportation ofFor , unirradiated fuel, nber of the Northeast Int tate Low-Level RadioactiveWa irradiated fuel, and 1he PSEG Site is Bamw 1, SC. The route was determined bythe radwaste at Alternative puter code to be 689 mi. for a commercial truck.Ro Ste -1 72 7-3, and 1 ,RADTRAN (Referer es 5.7-8 and 5.7-9) for truck

ship 41 idictes• that the . GIS (Reference 5. '-6) computer code. The population

dat. alternative sites are not • ) computer code is/based on the 2000 U.S. census. Allrad biusly superior to the •lby legal-weight tr s to the Bamwell, SC site over

The ~~EGS~t•f~• by multiplying t~e number of radwaste shipments per year

by the per-shipment doses. The nul bers of shipm Its per year are identified in

Table 5.7-7. The results for the rad• aste shipme t/ased on the RADTRAN (References 5.7-8

and 5.7-9) analyses are provided i• Table 5.7-11.

[ ] ConclusionA detailed analysis of the environz ;ntal impacts for the transportation of unirradiated fuel,

irradiated fuel, and radioactive wa' transporte .. ~........ .- ..... d inaccordance with 10 CFR 51 .52(b).' Re l c it 7 . per Y ear.0p er t

The new plant has sufficient fuel pool storage od of

Sfive years. At this timle, it iassumed that therll is sufficient storage capacity to permit irradiated

'incident-ie toaste fromnot

smallbetweennd thepository."uestion

f•

Replace with "5.7.2.4" per RAI t the requireme ts of shipping casks available at the time the fuelNo. nv Question_ 6-0-5- med all shipme ts are by truck. The shipping weight complies with

Question.6.0.... . gov shins as appropriate. The total number ofshipments for the bounding plant i 654.ipeyear r 0.2 per day which meets the Table S-4requirement of less than onical effects of incident-free conditions oftransport are summarized i Tabp ... "V. he radiological effects of accidents in transport areprovided in Section 7.4. The ned by these analyses represent the contribution ofsuch effects to the environmental osts of licensing the reactor.

The population doses to the trans 7ort crew and onlookers resulting from the new plant exceedTable S-4 values. Three key reas s for these higher population doses relative to Ta S-4 arethe shipping distances assumed fc r these analyses relative to the assumptions used ir WASH-1238 (Reference 5.7-10), the use of the maximum dose rate in the RADTRAN (Refere ices5.7-8 and 5.7-9) calculations, and be use of 30 minutes as the average time for a trucl stop inthe calculations

Replac withAdd "normalized to theReTabe 5.-4 preference reactor" per RAI Rev. 1:'RAb N n5.7-14 No. Re-. Q

,Question 6.0-5.,- A A A A AA A A A A A A A A A A A A A_ýý

Enclosure 2 Page 13

Repllace'with .....

PSEG Site corresponding values fromESP Application Insert E per RAI No.

Part 3, Environmental Report Env-12 Question 6.0-6.

Table 5.7-1 (Sheet I of 3)Uranium Fuel Cycle Datai=j

Reference LWR New PlantParameter Table S-3 Bounding Value

MWe 1000 2300Capacity Factor 0.8 00.99MWe (Net) 800 2-486Scale Factor~b 1.00 2-3Environmental ConsiderationsLand (Acres)

Temporarily committed() 100 -2W3Undisturbed area 79 -246Disturbed area 22 60Permanently committed 13 36Overburden moved (millions of MT) 2.8 -7.

Water (Millions of gallons)Discharged to air 160 437-Discharged to water bodies 11,090 30,290Discharged to ground 127 347-

Total 11,377Electrical energy (thousands of MW-hour) 323Equivalent coal (thousands of MT) 118 -322-(Estimated CO2, thousands of U.S. tons) (381) (404,0-Natural gas (millions of scf) 135 -369(Estimated CO2, thousands of U.S. tons) (8.25) (22-6)

Effluents-Chemical (MT)SO• 4400 42016NOIId) 1190 32-0Hydrocarbons 14 W3Co 29.6 8Particulates 1154 34-62F 0.67 4-83HCI 0.014 0038SO4- 9.9 2-7-4NOr- 25.8 J"Fluoride 12.9 3&2Ca'4 5.4 44CI- 8.5 234Na* 12.1 -3-30NH3 10.0 214Fe 0.4 44

Tailings solutions (thousands of MT) 240 &6%Solids 91,000 248,644Effluents-Radiological (Curies)

Rn-2220 Note (e) Note (e)Ra-226 0.02 0(&Th-230 0.02

Rev. 15.7-17

Enclosure 2 Page 14


Riepl'a'ce wifth co'rres'p'ond'in'g.values from Insert F per RAINo. Env-12 Question 6.0-6.

Prat 3, Environmental Report

Table 5.7-1 (Sheet 2 of 3)Uranium Fuel Cycle Datal*

Reference LWR New PlantParameter Table S-3 Bounding Value

Uranium 0.034 0.093Tritium (thousands) 18.1 494C-14 24 66Kr-85 (thousands) 400 -1093Ru-106 0.14 0-381-129 1.31-131 0.83 2.2T7Tc-99(e) Note (e) Note (e)

Fission products and TRU(f) 0.203 0.664Liquids

Uranium and daughters 2.1 &TRa-226 0.0034 J00093Th-230 0.0015 0.0941-Th-234 0.01

Fission and activation products 5.9E-06 ,.61 E 05Solids (buried on site):

Other than HLWf) (shallow) 11,300 30,863TRU(f) and HLW t ) (deep) 1.1E+07 -30E=9Effluents - thermal (billions of Btu) 4063 441097

Transportation (person-rem)Exposure of workers and general 2.5 68publicOccupational exposure 22.6 64-

Notes:

a) In some cases where no entry appears in NRC Table S-3 it is clear from the backgrounddocuments that the matter was addressed and that, in effect, the table should be read as if aspecific zero entry had been made. However, there are other areas that are not addressed at allin the table. NRC Table S-3 does not include health effects from the effluents described in thetable, or estimates of releases of radon-222 from the uranium fuel cycle or estimates oftechnebum-99 released from waste management or reprocessing activities. Radiological impactsof these two radionuclides are addressed in NUREG-1437, "Generic Environmental ImpactStatement for License Renewal of Nuclear Plants," dated May 1996, and it was concluded thatthe health effects from these two radionuclides posed a small risk.

Data supporting this table are given in the 'Environmental Survey of the Uranium Fuel Cycle,"WASH-1 248, April 1974; the "Environmental Survey of Reprocessing and Waste ManagementPortion of the LWR Fuel Cycle," NUREG-01 16 (Supp. 1 to WASH-1248); the 'Public Commentsand Task Force Responses Regarding the Environmental Survey of the Reprocessing and WasteManagement Portions of the LWR Fuel Cycle," NUREG-0216 (Supp.2 to WASH-1248); and in therecord of final rulemaking pertaining to Uranium Fuel Cycle Impacts from Spent Fuel

Rev. 15.7-18

Enclosure 2 Page 15

Referen e wonithcresodn

• ,--(values from Insert G per RAI No.ESPSEG..liction LEnv-12 Question 6-0-6.


Table 5.7'-3Total Effective Dose Equivalent from Uranium Mining and Milling

New PlantReference Bounding

Parameter LWR ValueMWe 1000 2300Capacity Factor 0.80MWe (Net) 800Scale Factor•= 1.00 243

Rn-222 (Ci~yr) 5191 4,474Tc-99 (Cityr) 0.012 0.033

100-yr Dose CommitmentRn-222 (person-rem) 140 382Tc-99 (person-rem) 100 W3Gaseous effluents excluding Rn-222 and 400 409reactor operation (person-rem)Liquid effluents excluding reactor 200operations (person-rem)Total 100 yr dose commitment (person-rem) 840 2293

a) The Scale Factor is the net MWe of the bounding new plant (Dual Unit APIO000) divided by thenet MWe of the reference LWR.

Rev. 15.7-21

Enclosure 2 Page 16



Table 5.7-410 CFR 51.52 Table S-4 Conditions and Reactor Technologies Attributes

10 CFR 51.52 10 CFR 51.52Condition Parameter Table S-4 ABWR APIODO U.S. EPR US-APWR

(a)(1)

(a)(2)

(a)(2)

(a)(2)

(a)(3)

(a)(3)

(a)(4)

Power Level

(MWt)

Fuel Form

3600 4300 3400 4590 4451

Sintered U0 2Pellets

4

Sirtered U0 2Pellets

4.9MaximumEnrichment

Clad

BumupOAWdIU)

Peak RedBurnup

Cooling Period

All radioar•,,ewaste is shipped

in solid form

Zircaloy Zircaloy-2

33,000 52,000

62,000

(a)(5) Shipment Mode(Unirrdiiated

Fuel, IrradatedFuel, Radioactive

Waste)

NtA ElectricGeneration

(MWe)

WA Annual FuelLoad (MU1)

WA RadwasteVolume

(ft3/r)

WA RadwasteActivityMBq/yr~(Cifyr)

90 Days

Al radioactivew.aste is

shipped insolid form

Truck

1000

35.0

108(3814)

5 years

AM radioactivewaste is


Truck

1500

44.7

165.0(5827)

Sintered U02Pellets

4.95

zirlo

48,700

62.001

5 years

All radioactivevraste is

shipped insolid fomr

Truck

1150

24A

55.6(1964)

Sintered U0 2Pellets

4.95

M5

54,000

62.000

5 years

All racioactivewaste is


Truck

1600

37.5

187.4(6618)

Sintered U02Pellets



Table 5.7-5New Fuel Shipment Data

ABWR APt000 U.S. EPR US-APWR

Annual Annual Annual Annual Annual Annual Annual Annual

Reload Number Reload Number Reload Number Reload NumberQuantity of Quantity Gf Ouantit of Quantity of

(MTU) Shipments (MTU) Shipments (MTU) Shipments (MTU) Shipmen

447 6.1 24A 3.8 37.5 7.5 35.0

.Reepla~ce with '15.3"[.per RAI No. Env-1 2:Question :6.0-5.

Rev. 15.7-23

Enclosure 2 Page 18



Table 5.7-8Annual Shipment Summary

ABWR APIO00 U.S. EPR US-APWR

Slipments Shipments Shipments Shipmentsperyear per year per year peryear

New Fuel 6-1 3.8 7.5 QSpent; Fuel 24.8 13.6 20.8 19.4

Radwaste .23.7 7._99 12Z• 21A8

:RAI No. Env-12:Question 6.0-5. Replace with "46.5" per

Rev. 15.7-26

Enclosure 2 Page 19



Table 5.7-9Unirradiated (New) Fuel Shipment Cumulative Dose

Expo-d PopUbin Dose per AMR APIooo U-. EMr US-APShisent

(]pez30-Sv) Dose per Doe per caser Dose perShpmts Year Shipment Year qipamft Year Shkpmerns Yearper Year (Persa--ýM per Year oepe~sn-se) per Year fperson-S-4 ~ l

Trznsportadmo wc*ers 1.44E435 8.1 BJES7-5 asB &.47E-05 7.5 I.OSE-04 -a &; 5

General PuL_

Transit 4.90E-08 &I 2.9DE-5 3-8 1.8SE.5 7.5 3.69E-05 -2a 4 j-OI

Stops 2.2E-05 SI 1.78E-04 IS IIE-04 7-5 -2__

tOW 3.41E-05 255.-04 1. Z5 E- --- - ,

Replace each value wth -5.3.peLRAI No. Ev-1'2 Question '6-0-5.

I

, Per RAI Env-12 Question 6.0-5, replace values with:

Transportation Workers - 7.63E-05Transit - 2-60E-05

:Stops - 1.55E-04Total - 1.81 E-G4

Rev. 15.7-27

Enclosure 2 Page 20

Insert A per RAI No. Env-12 Question 6.0-1:

5.7.2.1.12 Heat Load

In regards to the heat load, the US-APWR has the bounding value among theconsidered technologies. The heat load per irradiated fuel cask in transit for the US-APWR is 26,888 Btu/hr. This is less than the value of 250,000 Btu/hr given in Table S-4of 10 CFR 51.52. Therefore, the heat load generated by the US-APWR fuel per spentfuel cask will not result in significant environmental effects during transit.

Insert B (new subsection) per RAI No. Env-12 Question 6.0-5:

5.7.2.3 Comparison to 10 CFR 51.52 Table S-4

For an equal comparison to the reference reactor in 10 CFR 51.52 Table S-4, thenumber of shipments in Table 5.7-8 for each of the reactor technologies beingconsidered must be normalized. For each technology, the number of shipments isnormalized based on net electric generation (see Table 5.7-4) relative to the 1100 MWeand 80 percent capacity factor reference reactor analyzed in WASH-1238 (Reference5.7-10). Additionally, the unirradiated fuel shipments are adjusted to account for theinitial core loading in the annual number of shipments for each reactor technology. Thespent fuel shipments are scaled to reflect the capacity of 0.5 MTU/container used for thereference reactor. The radwaste shipments are scaled to reflect a capacity of 82.6ft3/shipment (2.34 m3/shipment) used for the reference reactor. The resulting annualtruck shipments normalized to the reference reactor are summarized in Table 5.7-13.

The incident-free consequences are determined based on the normalized number ofshipments for each reactor technology. The doses per shipment for unirradiated fuel,irradiated fuel, and radwaste are indicated in Tables 5.7-9, 5.7-10, and 5.7-11,respectively. The summary of the incident-free doses are shown in Table 5.7-14.

Insert C (new table) to the end of Section 5.7 per RAI No. Env-12 Question 6.0-5:

Table 5.7-13Annual Normalized Shipment Summary

Reference Normalized Normalized Normalized NormalizedShipments/year LWR ABWR AP1000 U.S. EPR US-APWR

New Fuel 6 4.3 1.8 4.9 3.5Spent Fuel 60 54.5 19.5 42.7 39.8Radwaste 46 15.8 3.4 7.5 13.5

Total 112 74.6 24.7 55.1 56.8

Enclosure 2 Page 21

Insert D (new table) to the end of Section 5.7 rer RAI No. Env-1 2 Question 6.0-5:

Table 5.7-14Total Normalized Shipment Cumulative Dose Summary

ABWR AP1000 U.S. EPR US-APWR10 CFR

51.52 Table Dose per Dose per Dose per Dose perExposed Population S-4 Year Year Year Year

pe

pseg early site permit application, response to request ...christine t. neely director regulatory...

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