t 7f AND A i 1EPORIT SAND84-1007 * Unlimited Release * UC-70Printed December 1985

Nevada Nuclear Waste Storage Investigations Project

The Potential Effect of WaterInflux on the Dissolution Rateof U02 in Spent Fuel at theYucca Mounitain, Nevada Site

Jeffrey W. Braithwaite

Prepared bySandia National LaboratoriesAlbuquerque, New Mexico 87185 and Livermore, California 94550for the United States Department of Energyunder Contract DE-AC04-76DP00789

0r0C

0C-Cmz

-Ti

3m

SF290oo(941)

,I

DistributionCategory UC-70

SANDB4-1007Unlimited Release

Printed December 1985

THE POTENTIAL EFFECT OF WATER INFLUX ON THE DISSOLUTIONRATE OF U02 IN SPENT FUEL AT THE YUCCA MOUNTAIN, NEVADA SITE

by

Jeffrey W. BraithwaiteNNWSI Repository Performance Assessments Division 6312

Sandia National LaboratoriesAlbuquerque, New Mexico 87185

ABSTRACT

This study identifies the potential effect of groundwater influx on

the dissolution rate of uranium dioxide (UO2) from spent fuel for

expected conditions at a prospective site for the disposal of radioactive

waste. The analysis is based on the hydrological characteristics of the

unsaturated zone at Yucca Mountain, Nevada. Although conditions that

could lead to inundation of the waste packages are improbable at the

site, dissolution resulting from either complete or partial exposure of

the spent fuel to water was considered. Estimates were made of the lower

limit of water-contact time with the UO2 in spent fuel under partially

saturated conditions. If spent fuel is inundated, the dissolution rate

of the UO2 is constrained by its solubility limit. If partially

saturated conditions exist, the uranium dioxide leach rate could control

the dissolution process. Under conditions of both partial or complete

saturation, the dissolution rate of the uranium is shown to be a linear

function of the water influx. The practical application of the results

of this study for determining radionuclide-release rates from spent fuel

is also discussed.

i

TABLE OF CONTENTS

Page

ABSTRACT . . . . . . . . . . . . . . . . . . . . . . . . . .LIST OF FIGURES . . . . . . . . . . . . . . . . . . . . . . . . .iiiLIST OF TABLES . . . . . . . . . . . . . . . . . . . . . . . . . iv

INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . 1

ASSUMPTIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . 2Temperature . . . . . . . . . . . . . . . . . . . . . . . . . 2Water Chemistry .... . . . . . . . . . . . . . . . . . . . 3Water Influx . . . . . . . . . . . . . . . . . . . . . . . . 3Water Contact Time .... . . . . . . . . . . . . . . . . . 7

DETERMINATION OF THE URANIUM DIOXIDE DISSOLUTION RATE ... . . . 10Candidate Equations .... . . . . . . . . . . . . . . . . . 10Dissolution Rate as a Function of Water Flux . . . . . . . . . 14Identification of Rate-Controlling Description . . . . . . . . 19Saturated Conditions . . . . . . . . . . . . . . . . . . . . 19Partially Saturated Conditions . . . . . . . . . . . . . . . 21

APPLICABILITY TO DETERMINING RADIONUCLIDE RELEASE RATES . . . . . 21

SUMMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25

REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . 26

APPENDIX .... . . . . . . . . . . . . . . . . . . . . . . . . 28References . . . . . . . . . . . . . . . . . . . . . . . . . . 33

ii

LIST OF FIGURES

Firure Page

1 Pictorial representation of how water can contact thespent fuel in either a partially saturated or saturatedenvironment. 6

2 Two potential modes of water collection . . . . . . . . . 8

3 Leach rate for selected elements from spent fuel in0.03 K sodium bicarbonate at 25-C . . . . . . . . . . . . 13

4 The effect of water flux on the dissolution rate of uraniumdioxide under saturated conditions for the two potentialemplacement configurations and the applicable rate-controlling dissolution descriptions . . . . . . . . . . . 16

5 The effect of water flux on the dissolution rate of uraniumdioxide under partialiy saturated conditions for thevertical emplacement configuration and the three rate-controlling dissolution descriptions being considered . . 17

6 The effect of water flux on the dissolution rate of uraniumdioxide under partially saturated conditions for thehorizontal emplacement configuration and the three rate-controlling dissolution descriptions being considered . . . 18

7 The dissolution rate of uranium dioxide from spent fuelunder saturated conditions and for both potentialemplacement configurations . . . . . . . . . . . . . . . . 20

8 The dissolution rate of uranium dioxide from spent fuelunder partially saturated conditions and for both potentialemplacement configurations . . . . . . . . . . . . . . . . 22

lii

LIST OF TABLES

Table Pate

1 Chemical composition of the groundwater for determiningsolubility limits and leach rates . . . . . . . . . . . . 4

2 Characteristics of waste packages and of water flowthrough them. These data are given as a function ofemplacement configuration and water collection mode . . . . 9

3 Characteristics of spent fuel and spent-fuel wastepackages . . . . . . . . . . . . . . . . . . . . . . . . . 15

iv

INTRODUCTION

As part of the Office of Civilian Radioactive Waste Management

program, a feasibility study for determining the potential of Yucca

Mountain, Nevada, as a repository site for high-level radioactive waste,

is being conducted under the direction of the Nevada Nuclear Waste

Storage Investigations (NNWSI) Project of the U.S. Department of Energy

(DOE). The objective of the program is to develop facilities, processes,

and systems for safe Handling and long-term containment of radioactive

waste, in compliance with state and federal regulations, environmental

standards, and technical requirements.

The task of assessing the postclosure performance of the entire waste

isolation system at Yucca Mountain is being performed by the INWSI

Repository Assessments Division of Sandia National Laboratories (SUL).

An important part of the tools used to perform this systems assessment is

a "source term." A source term is a mathematical description of the rate

of radionuclide release as a function of time from the waste packages to

the geologic setting. A preliminary source term was developed at SNL

during the early part of 1984 that could be used in the system

performance assessments required by several NNWSI-related activities that

were ongoing during that year. These activities included the

environmental assessment, data priority studies, and eventually the site

characterization plan. A portion of the source-term development study is

detailed in this report.

Most of the pre-1984 system analyses that had been performed for

VKWSI and other projects included source terms that were based on

1) radionuclide leach rates, with release generally taken either as an

instantaneous pulse or as a linear function of time (PiSford, 1982;

de~arsily et al., 1977), or more recently 2) radionuclide solubility

limits coupled with a realistic groundwater flow rate (Thompson et al.,

1984). Other studies concerned with predicting the release rate of

radionuclides from the waste form have noted the important effect that

radionuclide solubility limits and the influx of water can have (Kerrisk,

1984; Oversby, 1983; Oztunali et al., 1983). However, none of these

studies produced results that are directly applicable to the conditions

expected within Yucca Mountain. Given this situation, the initial

-1-

portion of the SNL source-term effort consisted of analyzing the effect

of water influx on the dissolution rate of the uranium dioxide matrix of

spent fuel under conditions that could reasonably be expected at the

Yucca Mountain site. Because dissolution of the uranium matrix controls

the release of several radionuclides (Oversby and McCright, 1984;

Vandergraaf, 1980), this initial study constituted the basis for the

formulation of the principal part of the source-term model used in the

early system performance assessments, as described by Braithwaite and

Tierney (Hunter, 1984). The results of this initial portion of the

system-source-term study are given in this report. Prior to discussing

the effect of water flux on the uranium dioxide dissolution rate, the

assumed spent-fuel environment is described. In the final section, the

applicability of the U02 dissolution results to radionuclide release is

discussed.

ASSUMPTIONS

In lieu of sufficient data on hydrological, geochemical and

subsurface conditions, a number of very important assumptions were made

concerning the waste package environment. These assumptions are

concerned with the temperature of the spent fuel, the chemistry of the

groundwater, the flow of water into the emplacement borehole, and the

contact time between the spent fuel and the water. Each of these

subjects is addressed separately.

Temperature

The temperature of the spent fuel is always 250C. This is a

reasonable assumption for the following reasons:

o liquid water cannot contact the spent fuel until the spent fuel

temperature has dropped below the boiling point of the water and only

when the protective canister, as well as the Zircaloy fuel cladding,

has been breached. Recent thermal analyses indicate that spent fuel

will cool to boiling-point levels after approximately 1000 years

(Stein et al., 1984). However, studies characterizing the corrosion

-2-

behavior of proposed canister materials show a high potential for

containment times greater than 10,000 years in the Yucca Mountain

environment (Oversby and McCright, 1984). Because of this long-term

containment period, and the additional protection due to the Zircaloy

cladding (Rotbman, 1984), the waste would cool to near ambient

temperatures before contact with liquid water could be possible.

o an increase in temperature either decreases or has a minimal effect

on the solubility limit of uranium dioxide (Brush, 1979; Fullam,

1981).

o the activation energy for the leaching rate of uranium dioxide can be

either positive or negative, depending upon the composition of the

groundwater. In either case the absolute value of the activation

energy is small--approximately 20 kJ/M (Thomas, 1983; Johnson et al.,

1981).

Water Chemistry

The groundwater maintains equilibration with air at a partial

pressure of 0.2 atm oxygen. This condition produces an oxidizing

electrochemical potential in the groundwater of approximately 700 mV.

The groundwater composition used for determining the solubility limit and

applicable leach-rate calculations is as shown in Table 1. The source of

this analysis was the water sample obtained from a nearby well (denoted

J-13) located at Jackass Flats adjacent to Yucca Mountain.

Water Influx

Water flow through the unsaturated zone of Yucca Mountain should be

limited to that which corresponds to downward flux. Recent studies

(Sinnock et al., 1984; Peters et al., 1984; Braithwaite and Nimick, 1984)

of water flow through the unsaturated zone at Yucca Mountain indicate

that:

o water percolation rates are low. A flux of 0.1 mm/yr or less is

consistent with actual pressure and saturation measurements, and an

-3-

TABLI 1. Chemical composition of the groundwater used fordetermining solubility limits and leach rates.

Component Concentration (ppm)

HCO3 143

Si02 64

Na+

so4

Ca+2

18

12

No03 10

Cl- 6.4

5.5

F- 2.1

1.7

42 0.1

Fe 0.044

V 0.043

Sr 0.04

Al 0.026

Ba 0.0023

Hn 0.0011

Source: Kerrisk, 1984, based on water sample from well J-13, JackassFlats, NV.

-4-

upper bound of 1 m1/yr is probably justified. The saturated

hydraulic conductivity of the fractured host rock is probably very

high (> 1000 mm/yr).

o at these percolation rates, saturation is less than 1.0, and the

suction heads are sufficiently high that water movement will take

place only in the matrix of this highly fractured rock.

This information provides an estimate of the probable path and

quantity of water flow in and out of the emplacement horizon, but does

not assess the potential for water movement through the waste package or

through the thermally and mechanically disturbed host rock surrounding

the emplacement borehole. Because all emplacement configurations contain

an air gap between the waste package and the borehole wall, the results

given in Fernandez and Freshley (1984) can be used to predict that none

of the percolating water will actually contact the packages. This

situation leads to a probable aqueous release of near zero. However,

because release calculations must be performed to assess the performance

of the geologic setting, the following water influx/characteristics were

postulated for this study:

o both saturated (inundated) and partially saturated conditions were

considered. Because the permeability of the host rock mass is much

higher than any projected percolation rate, conditions that would

allow the waste package to become inundated would not be expected.

However, during the first several thousand years, it is conceivable

that a breech could occur in the top of a waste canister that would

allow the inside of the canister to fill with water (see Figure 1).

Another reason for including the saturated environment was to allow

the entire spent-fuel surface to be exposed to water. Because leach

rates are usually dependent on the amount of wetted surface area,

this assumption allows any surface-area effect to be bounded.

o the rate of water influx was based on the method of emplacement. Two

configurations of waste package emplacement, vertical and horizontal,

-5-

Partially Saturated Inundated

L////J//At_ \ , -_ Bz _

V d

419IA

eF

6141

Sf

661

V -- I

0i*66 6 A 4EO6 6 6

EXPOSED SPENT FUEL RODS

Figure 1. Pictorial representation of how water can contact the spent fuel ineither a partially saturated or saturated environment. The figure onthe left shows water droplets flowing over exposed U02 pellets. Thefigure on the right shows how an early breech in a waste container couldlead to inundated fuel rods.

are being considered for use at Yucca Mountain. The method of

emplacement has an important effect on the amount of water that

potentially could come in contact with each waste package for the

following reason:

the quantity of water influx was calculated assuming two

potential sources of water accumulation: 1) any percolating

water intersecting a mined area (including drifts and

emplacement boreholes) is collected and dispersed equally to all

vertically emplaced packages and 2) only the percolating water

that directly intersects the emplacement borehole contacts the

package (see Fig. 2). The first mode was used only for vertical

emplacement because horizontal boreholes will be constructed

above the-floor of all access and emplacement drifts and the

second mode is applicable to both horizontal and vertical

emplacement methods.

o all water that flows into the emplacement borehole contacts exposed

UO 2 The influence of claddings or any other waste packaging

components designed to protect the fuel and prevent exposure to water

was not considered in this study (except as previously noted to delay

the onset of dissolution until lower temperatures exist). Also, it

is expected that a substantial fraction of the water would not

contact the spent fuel because of the 60 to 70 percent void volume

that exists inside an emplacement borehole.

o a constant and uniform distribution of water flow and flux was

assumed across the entire emplacement horizon.

Water Contact Time

A relatively simple analysis was performed to determine lower limits

on the time that water can remain in contact with the UO2 surface

(contact time). A description of the analysis for the partially

saturated condition is given in the Appendix and a summary of the

important results for both partially saturated and the unexpected,

short-term saturated conditions is given in Table 2.

-7-

I I I I I I I 1%.. PERCOLATINOII I I I I I I WATER

I ~~~I I I I

m I III II I

f II

WASTE PACKAGE

MODE I

IHORIZONTAL EMPLACEMENT

VERTICAL EMPLACEMENT

MODE 2

Figure 2. Two potential modes of water collection.

NOTE: In Mode 1, all the water that would normally contact amined area (including access drifts and emplacement boreholes)is collected and evenly distributed to the waste packages.This mode is applicable only to vertical emplacement. In Mode2, only the water that contacts the emplacement borehole crosssection is available to contact the waste packages.

-8-

TABLE 2. Characteristics of waste packages and of water flow throughthem. These data are given as a function of emplacementconfiguration method and water collection mode.*

Horizontal Emplacement Vertical EmplacementFlow Hode 2 Flow Mode I Flow Mode 2

Waste package characteristics:

Borehole diameter (cm) 69 61 61Package length (cm) 450 450 450Borehole free volume (.) 70 62 62Volumetric flux 3.1 50 0.3

(liters/yr/package)

Water flow characteristics:

Inundated conditions:

Mean water retention time 380 16 2800(yr)

Partially saturated conditions (see Appendix):

Droplet flow rate 4 x 10-3 0.07 4 x 10-4(number/s/pkg)

Droplet retention time (s) 2.3 8.8 8.8

Surface area covered 1.6 x 10-7 1 x 10-5 6 x 10-8(m 2 /package)

Percentage surface area 1 x 10-7 7 x 10-6 4 x 10-8covered

C: constant in equation l 5 x 10-5 2 x 10-4 2 x 10-4(yr * m2/m3)

*NOTE: For an explanation of the two water modes, refer to the text orFig 2. All of the flow rates and area coverages shown are per mm/yr of waterflux. The water retention time for the inundated conditions was calculatedusing the information contained in the first four lines of this table.

-9-

Based on a mass balance, the surface-area coverage at any time for

partially saturated conditions is given by the following equation:

SA . C * F * A, (1)

where

SA - surface area covered (a of UO2/package)

F - flux of percolating water (m3/m2/yr)

A = effective horizontal intercept area of groundwater by the

package (m2)

C - factor accounting for the contact time of the droplets and

their ratio of contact area to volume (yr * m2 /m3)(see

Table 2 for actual values).

DETERMINATION OF THE URANIUM DIOXIDE DISSOLUTION RATE

Candidate Equations

The dissolution or leach rate of the uranium dioxide matrix

represents the kinetics of the dissolution process and is a function of

the aqueous environment and the chemical and physical properties of the

spent fuel. Leach rates are typically determined experimentally in

aqueous solution replenished often enough to ensure that any

rate-retarding effects of soluble products are not important. However,

if the leach rate is fast and little water is available, the

concentration of the solubilized product can increase to its solubility

limit. At this point, no further net dissolution will occur. Thus,

saturation of the available water with a dissolved product establishes an

upper bound on the dissolution rate.

The low level of water influx and the unsaturated conditions

expected at Yucca Mountain appear suitable to allow either solubility

limits or leach rates to control the dissolution rate. That is, if water

comes in contact with the waste for relatively long times (e.g., in a

water-filled container with breached areas only in the top), solubility

limits coupled with water influx could control the rate. On the other

hand, if the water-contact time is small (a.g., droplet flow over

-10-

individual spent-fuel rods under partially saturated conditions), leach

rates may control the dissolution. To determine which of these two

descriptions is appropriate for uranium dioxide dissolution under the

assumed conditions, the analysis detailed in the remainder of this

section was performed.

Because of the half life of uranium (4.5 x 109 years), radioactive

decay will not significantly affect the quantity of uranium available for

leaching over reasonable time periods. For convenience in this analysis,

the rate of dissolution of the uranium matrix is defined as the mass

fraction of uranium dissolved per year. This definition is related to

the curie release-rate.expression used by the NRC in 10 CFR 60 when the

total available mass is taken to be the inventory in the waste package

1000 years after closure. The mass-fraction-based release rate for the

uranium dioxide matrix is-given by equations (2) through (4) below.

For the dissolution rate controlled by the solubility limit and water

influx:

Rm =F A *.S/H , (2)

where

R = fractional release rate of the uranium dioxide matrix basedm

on mass (1/year)

Sm = solubility limit of the uranium dioxide matrix (kg/m )

im = mass of uranium contained in the waste package at the time of

closure (kg)

F = flux of percolating water (m 3/m2/yr)

A -effective horizontal intercept area of the groundwater by the

package (m2).

Equation (2) is applicable to both unsaturated and partially saturated

conditions because if all water flowing through the repository becomes

saturated, the manner in which water contacts the uranium is irrelevant.

The dissolution rate controlled by the leach rate is expressed as:

Rm = TA * SA/Mm (3)

-11-

where

LR = leach rate of the uranium matrix (kg/m /yr).

Equation (3) is directly applicable to the constantly wetted surface area

or saturated conditions under which the leach-rate measurements are made

(in this case SA - geometric surface area). Additionally, equation (1)

can be substituted into equation (3) to give an expression defining

release rate under the partially saturated conditions in which exposed

surface area may not be constant:

R = C * LB * A * F./ (4)m mn

To determine which of these two equations best describe the rate of

dissolution under the postulated conditions, site-specific information is

required concerning waste-package design, repository design, the uranium

dioxide leach rate, and the uranium dioxide solubility limit. Several

researchers have studied the dissolution characteristics of U02 in both

deionized water and carbonate-containing solutions (Wilson and Oversby,

1984; Katayama et al., 1980; Thomas, 1983; Johnson et al., 1981; Hiskey,

1979; Grandstaff, 1976). gone of these pre-1984 experimental studies

specifically used the chemical composition of the groundwater (e.g.,

150 ppm bicarbonate) or the partially saturated conditions expected at

Yucca Mountain. An initial uranium leach rate of 0.4 kg/m /yr and a

steady-state leach rate of 7 x 10 kg//m /yr were chosen from the

results shown in Fig. 3 as the most closely representative for the

expected environment. The initial leach rate is the dissolution rate

measured within the first few days after exposing fresh, unleached

uranium dioxide to a solution. The steady-state leach rate is the

dissolution rate measured after an extended exposure period (typically

greater than 1 year) by which time the rate-inhibiting effects of

alteration-product layers have become important. For more detail on the

effect of alteration products, see the last section of this report. The

initial leach rate was included in the calculations because it represents

the fastest dissolution possible in a given solution. The steady-state

-12-

ty5P'

I laTIME. days

Figure 3. Leach rate of selected elements from spent fuel in 0.03 M sodiumbicarbonate solution at 25-C.

NOTE: The leach rate given is equal to the fraction of eachelement extracted divided by the leach time interval (days) and thespecific surface area (cm2/g) of the spent fuel sample. Thisfigure was reproduced from Katayama et al. (1980).

-13-

data were included because they represent a lower bound on the dissolution

rate in solutions in which soluble products are not allowed to concentrate.

In addition, when considering release from spent fuel, Vandergraaf (1980)

concluded that leach tests must be conducted for a minimum of 300 days before

results can be regarded as meaningful. The leach-rate data given in Fig. 3

were selected because they represent the longest term data available for spent

fuel dissolution in an environment containing carbonate and because they

appear to be consistent with those identified by the other researchers. These

rates may be conservatively high (by up to a factor of 4) because, as Hiskey

(1979) noted, the uranium dioxide dissolution rate is proportional to the

square root of the carbonate concentration and Katayama's leaching solution

contained an order of magnitude higher carbonate than Yucca Mountain water

probably does.

Solubility limits for most of the important radionuclides were calculated

by Kerrisk (1984) using the geochemical code EQ3/6. For this study, only the

total solubility limit for uranium is important (0.05 kg/m ). summaries of

relevant repository layout and waste package design information are given in

Tables 2 and 3.

Dissolution Rate as a Function of Water Flux

Equations 2, 3, and 4 can be used to calculate the uranium dissolution

rate for the two rate-controlling processes being considered. As mentioned

earlier, the flux of water through the repository region has not yet been

accurately determined. Because this parameter represents the principal

unknown in both equation 2 (solubility controlled) and equation 4

(leach-rate-controlled under partially saturated conditions), the uranium

dissolution rate has been calculated and plotted in Figs. 4, 5, and 6 as a

function of water flux for the two emplacement configurations and

water-collection conditions assumed in this study. Fig. 4 represents the

response under saturated conditions for vertical and horizontal emplacement

configurations; Fig. 5 is for partially saturated conditions and vertical

emplacement, and Fig. 6 is for partially saturated conditions and horizontal

emplacement. The notation of flow mode refers to the assumption concerning

the disposition of infiltrating water that would contact a mined region (see

Fig. 2). Mode 1 yields conservatively high values for water influx, but as

explained in the assumptions section, it is

_14-

TABLE 3. Characteristics of spent fuel and spent-fuel wastepackages.

SPENT FUEL

Pin length

activetotal

Pin diameter

Pellet diameter

Cladding thickness

366 cm385 cm

0.95 cm

0.82 cm

0.057 cm

WASTE PACKAGE

Thermal power

Number/acre at57 kW/acre

Outside diameter

Length

Total weight

lumber of pins

Weight of spent-fuel components:

UraniumMetal oxideZirconiumOther hardware

Total geometric surface area of U02

3.05 kW

18.7

50 cm

450 cm

4500 kg

1584

2766314475696

149

kgkgkgkg

Source: (Bechtel, 1979; Gregg and O'Neal, 1983)

-15-

1=..

0)QasC0

CD

M._

o..'.."

0I

f-.z0

-

n0U)U)

10-5

10-6

0sI

10-8

10-9

10-10

10-11

101-2 L10-2 lo-' 100 101 102

WATER FLUX (mm/yr)

FViure 4. The effect of water flux on the dissolution rate of uranium dioxideunder saturated conditions for the two potential emplacementconfigurations and the applicable rate-controlling dissolutiondesciv.tions.

10-4

haAs

0.0Ma-

.I0

3

crz0

J0C,)U)S

10-5 :r

1 0 -6.

10-7.

10-9 ..

10-11

10-12 -

10-2

I0-j

10-1 100 10 102 103

WATER FLUX (mm/yr)

Figure 5. The effect of water flux on the dissolution rate of uranium dioxideunder partially saturated conditions for the vertical emplacementconfiguration and the three rate-controlling dissolutiondescriptions being considered.

10-4

b-

b-

W

03C0

L)

I-

V)%0

V,)

Qcc)a

10-6

10-8

10-10

10-11

Ia

10-12 L10-2 10-1 100 101 102 1o3

WATER FLUX (mm/yr)

Figure 6. The effect of water flux on the dissolution rate of uranium dioxideunder partially saturated conditions for the horizontal emplacementconfiguration and the three rate-controlling dissolutiondescriptions being considered.

applicable only to vertical emplacement because horizontal boreholes are

located above drift floors. Mode 2 represents a more realistic upper-bound

condition. Because mode 2 collection yields much lower volumes of water that

can contact the fuel, it is not surprising that solubility-based dissolution

rates, which are a linear function of water flux, are also much lower than for

mode 1 collection.

As shown in Table 2, the mean contact time of the water for saturated

conditions would be very long (16 to 2800 years). With such slow water

movement, initial leach rates that occur for the first year of exposure,

should not be important. Therefore, in Fig. 4, only steady-state leach rates

were included. On the other hand, under partially saturated conditions, both

initial and steady-state leach rates could be important. With a predicted

droplet contact time of less than 10 seconds (see Table 2), initial leach

rates would be expected to dominate until the passivating alteration product

layer had been formed. After this time, steady-state leach rates could

control the dissolution rate.

Identification of Rate-Controllinx Description

The processes that control the dissolution rate for both emplacement

methods and both water-collection modes can be identified by considering the

results shown in Figs. 4, 5, and 6. The process that produces the slowest

dissolution rate is rate controlling. That is, a solubility limit cannot be

achieved if the leach rates are too slow and, as mentioned previously, the

dissolution rate cannot cause the solibility liimit to be exceeded at any leach

rate. The justification for the controlling-process selection is given

separately for saturated and partially saturated conditions in the following

paragraphs.

Saturated Conditions. As shown in Fig. 4, dissolution based on the

solubility limit constrains the rate until unrealistically high water flux

values are reached. At this point, the steady-state leach rate, which is not

a function of the flux under these inundated conditions, begins to dominate.

Therefore, if the canister interior or emplacement borehole becomes inundated,

the uranium solubility limit should constrain the dissolution rate of the

uranium dioxide. The appropriate data are replotted in Fig. 7 in which the

effect of water flux on the controlled dissolution rate of uranium dioxide

-19-

I-

b-0)a.C0

0-IL)

nI-

'C0~

-J%0

(I)3

I

0

10-2 10-1 100 10l 102 1o 3

WATER FLUX (mm/yr)

Figure 7. The dissolution rate of uranium dioxide froa spent fuel as afunction of water flux under saturated conditions and for bothpotential emplacement configurations. The responses shown in thisfigure are based on the rate-controlling process descriptionsapplicable for these conditions. The cross-hatched area Indicatesthe currently expected range of water-flux values at the Yucca

Mountain repository horizon.

under saturated conditions is shown. The crosshatched area on this

figure indicates the water-flux range that is consistent with

hydrological calculations and field measurements at the projected horizon

for the repository at Yucca Mountain. A value of flux greater than

1 mm/yr is unlikely according to those calculations and measurements,

which are, however, consistent or lower than values of flux equal to

0.01 mm/yr. Dissolution rates at a water flux of up to 1000 mms/yr were

included in these calculations to allow the dissolution rate to be

plotted under partially saturated conditions and steady-state leach-rate

control (Fig. 5).

Partially Saturated Conditions. As shown in Figs. 5 and 6, leach

rates should control the dissolution of uranium. This results from the

very short contact times that were calculated for the partially saturated

conditions. Whereas one expects the steady-state leach rates to dominate

over the long term, the initial leach rate was selected for the remaining

calculations and discussion because of the uncertainty in droplet flow

time. The travel time for the droplet to flow over the uranium dioxide

that was used, less than 10 seconds, represents the minimum possible

value. Surface contamination, rod-to-rod contact, and the presence of

discrete joints between individual pellets could dramatically increase

droplet contact time and therefore the overall rate of dissolution.

Again, the appropriately chosen data are replotted in Fig. 8 showing the

effect of water flux on the controlled dissolution of uranium dioxide

under partially saturated conditions. The observation that a very high

leach rate can still control the dissolution rate is important because it

means leach rates must be accurately determined if realistically

appropriate source terms are to be formulated.

APPLICABILITY TO DETERMINING RADIONUCLIDE RELEASE RATES

The dissolution rate of the uranium dioxide matrix of spent fuel-

plays an important role in determining the release rate of a number of

fission product and actinide radionuclides that are physically separated

from a leaching groundwater by this relatively impermeable matrix. For

-21-

he.

:0

A-

a.Cn0-

s

co

-

O)0)

cc

0

J0CoU)

nQo

I3

IJ

10-2 lo-' 100 lo 102 103

WATER FLUX (mm/vr)

vigure B. The dissolution rate of uranium dioxide from spent fuel as afunction of water flux under partially saturated conditions and forboth potential emplacement configurations. The responses shown inthis figure are based on the rate-controlling process descriptionsapplicable for these conditions. The cross-hatched area indicatesthe currently expected range of water-flux values at the YuccaMountain Repository horizon.

several radionuclides that are not contained within the matrix (e.g.,

portions of cesium, technetium, carbon, and iodine), the dissolution of

the matrix is unimportant.

In this subsection, information is presented to show why a modified

congruent-alteration mechanism is currently a reasonable approach that

would allow the release rate for the appropriate radionuclides to be

determined. In this discussion, congruent alteration simply means that

the availability or exposure of radionuclides to the aqueous environment

is limited by the extent of aqueous alteration of the uranium dioxide

matrix. As a portion of the matrix reacts with the groundwater, all

species contained within that portion are also subject to contact with

the groundwater. In addition, it can be assumed that once the matrix is

altered, all available species within that element, including uranium,

are solubilized. Thus, no solid layer of alteration products forms and,

with a uniform distribution of radionuclides throughout the spent fuel,

the fractional rate of dissolution of the other species is restricted to

the fractional rate of alteration of the uranium matrix. This mechanism

has been termed "congruent alteration" to distinguish it from the more

frequently discussed "congruent leaching" mechanism. Whereas most

characterization studies of spent-fuel dissolution support the conclusion

that species contained in the unleashed portions of the matrix undergo

congruent alteration with the matrix, there exists considerable confusion

as to whether these same species are actually solubilized in a congruent

manner (congruent leaching). Congruent alteration with complete product

solubilization is a conservative form of congruent leaching, and the

assumption that it controls the release of several radionuclides is

currently justified for the following reasons:

a) The presence of a rind layer consisting of solid reaction products

on the surface of dissolving spent fuel (Thomas, 1983) can play an

important role by forcing experimentally measured (or apparent)

individual release rates to deviate from those predicted simply by

matrix dissolution. This deviation occurs primarily for two

reasons: 1) A precipitation effect: Primarily because reactants

diffuse inward and soluble products diffuse outward through this

rind layer, the aqueous chemistry at the reacting interface is

different from the chemistry in the external groundwater.-23-

This chemical difference allows solid hydrated oxy-uranium products

of varying composition to precipitate as this rind layer. This

precipitation process selectively removes solubilized radionuclides,

and 2) A passivation effect: Oxygen has been shown to be a key

reactant in natural groundwaters because the leach rate of uranium

dioxide is dependent on the dissolved oxygen concentration (Hiskey,

1979; Grandstaff, 1976; Johnson, 1981). Therefore, for the

continued dissolution of uranium to occur, oxygen must diffuse

through the rind layer to the unreacted core interface. This

diffusion process causes the oxygen concentration at the interface

to be lower than in the bulk solution. The passivating effect (#2)

and the variable precipitation effect (#1) could account for a major

portion of both the decrease and the spread observed experimentally

in spent-fuel leach rates (see Fig. 3). By assuming congruent

alteration with soluble products, the inhibiting effects of oxygen

diffusion and solid-product formation are ignored and all species

are solubilized at the maximum rate possible.

b) Some oxidation of the uranium dioxide matrix, along with segregation

of many nonuranium species, occurs during irradiation and

preparation of experimental samples (Katayama et.al, 1980;

Vandergraaf, 1980). These processes, in conjunction with a newly

forming alteration layer, require that care be exercised when

applying short-term experimental dissolution data to the formulation

of conclusions concerning congruent alteration. To illustrate this

point, only 0.075% of a typical -2 + 4 mm fuel fragment (i.e., a

fragment with an equivalent spherical diameter between 2 and 4 mm)

would dissolve each year into large excesses of deionized water.

This rate corresponds to only 1 um/yr of penetration. With such

slow reaction rates, short-term data generally reflect the leaching

characteristics of the exposed, potentially perturbed surface and

not any bulk material.

-24-

c) If the composition and thickness of the alteration product layer

reaches steady-state values, then individual radionuclide release

rates should follow a congruent-leaching mechanism. A Canadian

researcher (Thomas, 1983) concluded that this is indeed the case in

the long term. Furthermore, the spent-fuel leach data shown in

Fig. 3 for periods greater than 1 year, generally shows a trend

toward congruent leaching. Finally, the conclusion was reached in a

review conducted for the VRC that the actinide products dissolve

congruently (Dayal et al., 1982). The actinides may be the primary

concern with an expected containment time of 1,000 to 10,000 years.

-SUMMARY

The effect of water flux or percolation rate on the dissolution of

uranium dioxide from spent fuel was investigated for conditions that

could be encountered within a waste package emplaced at Yucca Mountain.

Despite the low probability that spent fuel would be exposed to water,

allowances were made in this study for the occurrence of a breech in the

waste container and for partial or complete saturation. The most

significant results of this study indicate that:

1. If the spent fuel is saturated, the dissolution rate of the uranium

dioxide will be constrained by its solubility limit and the influx

of water.

2. If the spent fuel is exposed only to partially saturated conditions,

as is currently expected, then the leach rate of the uranium dioxide

should control its rate of dissolution.

3. The release rate of many nonuranium radionuclides can reliably be

predicted, using a congruent matrix alteration mechanism with

complete product solubilization.

-25-

REFERENCES

Bechtel National, Inc., "An Assessment of LWR Spent Fuel DisposalOptions," ONWI-39, Battelle Memorial Institute, Vol 3, July 1979, p.A-14.

Braithwaite, J. W., and F. B. Nimick, "Effect of Host-Rock Dissolution andPrecipitation on Permeability in a Nuclear Waste Repository in Tuff,"SAND84-0192, Sandia National Laboratories, September 1984, p. 35-39.

Brush, L. H, "The Solubility of Some Phases in the SystemU03-Na2 O-H20 in Aqueous Solutions at 60 and 900C," Ph.D.Dissertation, Harvard University, June 1980, p. 155.

deHarsily, G., et al., "Nuclear Waste Disposal: Can the GeologistGuarantee Isolation?," Science, 197 (4303), August 5, 1977, p. 520.

Dayal, R., at al., Nuclear Waste Management Technical Support in theDevelopment of Nuclear Waste Form Criteria for the NRC,"NUREG/CR-2333, U.S. Nuclear Regulatory Commission, February 1982,p. 18-39.

Fernandez, J. A., and H. D. Freshley, "Repository Sealing Concepts forthe Nevada Nuclear Waste Storage Investigations Project,"SAND83-1778, Sandia National Laboratories, August 1984, p. 33.

Fullam, H. T., "Solubility Effects in Waste-Glass/Demineralized-WaterSystems," PNL-3614, Pacific Northwest Laboratories, June 1981,p. 30-32.

Grandstaff, D. E., "A Kinetic Study of the Dissolution of Uraninits,"Econ. Geol., 71 (8), 1976, pp. 1495-1497, 1500.

Gregg, D., and W. O'Neal, "Initial Specifications for Nuclear WastePackage External Dimensions and materials," UCID-19926, LawrenceLivermore National Laboratories, September 1983.

Hiskey, J. B., "Kinetics of Uranium Dioxide Dissolution in AmmoniumCarbonate," Trans C Inst Min Metal, 88, 1979, p. 146-148.

Hunter, T. 0., and A. B. Muller, "Proceedings of the Workshop on theSource Term for Radionuclide Migration for High-Level Waste or SpentNuclear Fuel Under Realistic Repository Conditions," SAND85-0380,Sandia National Laboratories, July 1985, pp. 71-75.

Johnson, L. H., et al., "The Dissolution of Unirradiated Uranium DioxideFuel Under Hydrothermal Oxidizing Conditions," AECL-TR128, AtomicEnergy of Canada, LTD. April 1981, pp. 1, 2, 8, 10.

Katayama, Y. B., et al.,"Status Report on LWR Spent Fuel IAEA LeachTests," PNL-3173, Pacific Northwest Laboratories, March 1980, pp.3-9, 28.

-26-

Kerrisk, J. F., "Solubility Limits on Radionuclide Dissolution at a YuccaMountain Repository," LA-9995-MS, Log Alamos National Laboratory, May1984, pp. 1, 5, 32.

Oversby, V. M., "Performance Testing of Waste Forms in a TuffEnvironment," UCRL-90045, Lawrence Livermore National Laboratory,November 1983, p. 10.

Oversby, V. M., and R. D. McCright, "Laboratory Experiments Designed toProvide Limits on the Radionuclide Source Term for the NNWSIProject," UCRL-91257, Lawrence Livermore National Laboratory,November 1984, p. 6.

Oztunali, 0. I., et al., "Influence of Leach Rate and Other Parameters onGroundwater Migration," NUREG/CR-3130, Nuclear Regulatory Conmiission,February 1983.

Peters, R. R., et al., "Fracture and Matrix Hydrologic Characteristics ofTuffaceous Materials from Yucca Mountain, Nye County, Nevada,"SAND84-1471, Sandia National Laboratories, December 1984, p. 58-60.

Pigford, T. H., "Geological Disposal of Radioactive Waste," ChemicalEngineering Progress (CEP), pp. 18-26, March 1982, p. 18-19.

Rothman, A. J., "Potential Corrosion and Degradation Mechanisms ofZircaloy Cladding on Spent Nuclear Fuel in a Tuff Repository," UCID20172, Lawrence Livermore National Laboratory, September 1984, p. 40.

Sinnock, S., et al., "Preliminary Bounds on the Expected PreclosurePerformance of the Yucca Mountain Repository Site, Southern Nevada,"SAND84-1492, Sandia National Laboratories, December 1984, p. 13-14.

Stein, W., et al., "Thermal Analysis of NNWSI Conceptual Waste PackageDesigns," UCID-20091, Lawrence Livermore National Laboratory, April1984, p. 66-70.

Thomas, G. F., "The Dissolution of Unirradiated Uranium Dioxide FuelPellets," AECL-TR234, Atomic Energy of Canada, LTD, November 1983,pp. 2, 3, 7, 10-13.

Thompson, F. L., F. H. Dove, and K. M. Krupka, "Preliminary Upper-BoundConsequence Analysis for a Waste Repository at Yucca Mountain,Nevada," SAND 83-7475, Sandia National Laboratories, August 1984,p. 13.

Vandergraaf, T. T., "Leaching of Irradiated Uranium Dioxide Fuel,"AECL-TR100, Atomic Energy of Canada, LTD, May 1980, p. 13.

Wilson, C. N., and V. M. Overgby, "Spent-Fuel Cladding Containment CreditTests," UCRL 89869, Lawrence Livermore National Laboratory, September1984, p. 6-8.

-27-

APPENDIX

CHARACTERISTICS OF WATER FLOW OVER URANIUM DIOXIDE IN SPENT FUEL

IN AN UNSATURATED ENVIRONMENT

The analysis that was performed to determine the flow characteristics

of water over spent-fuel rods in an unsaturated environment is detailed

in this section. This initial study, although relatively simple in

scope, was performed so that any leach-rate affects on the dissolution of

spent fuel could be included in this report. The author is grateful to

Tom Hinkebein of Division 6257 of Sandia National Laboratories for his

valuable discussions concerning this analysis.

The primary assumption used in this analysis is that all water that

enters the emplacement borehole contacts bare spent fuel. As such, the

protective metal barriers (the container and Zircaloy cladding) have

corroded away, exposing the spent fuel with sufficient cladding remaining

only to support the spent-fuel pellets in an intact rod configuration.

With the very low water percolation rates expected through the

repository horizon (less than 50 1/yr/package) and the high geometric

surface areas per package (150 m 2), water would be expected to flow

over the spent fuel either in very thin films or possibly

intermittently. Levich (1962) points out that if a liquid film becomes

thin enough, the action of capillary forces will cause it to break up

into droplets. The velocity of a thin film flowing down a vertical

surface can be calculated using the following equation (Levich, 1962;

Szekely and Themelis, 1971):

K2V =LI (1)3',

where

V = velocity of the film over its horizontal cross section

h . film thickness

v = kinematic viscosity of water.

-28-

Performing a mass balance for a horizontal emplacement configuration,

complete wetting, and a percolation rate of 1 mm/yr, the steady-state

average velocity of the film must equal

1.21 x 10 cm/sec (2)

Equating equations 1 and 2 and solving for h results in an average film

thickness, under these given conditions, of only 5 micrometers. Since

this thickness is several orders of magnitude less-than the equivalent

hydrodynamic boundary layer thickness (the thickness at which the liquid

gains 10% of the bulk fluid velocity) for similar flow conditions around

rotating disks (Sohn and Wadsworth, 1979), it is unreasonable to expect

the film to move as a sheet. Thus, the flow should occur in discrete

droplets.

In order to identify both the size and the velocity of droplets

flowing down the surface of a spent-fuel rod, it is assumed that the

droplets remain approximately hemispherical and that a difference exists

between the advancing and receding contact angles in the droplets.

Adamson (1960) notes that this difference in contact angle has been

observed for water flow on many mineral surfaces. Based on data

contained in the Adamson reference, an advancing angle of 65 degrees and

a receding angle of 30 degrees *ere used.

For an approximately hemispherically shaped droplet flowing down a

vertical surface, the dynamic force balance is expressed as

Au dV =--r2i dV + vry(cose - cosO ) + mg (3)dt dy a r

-29-

where

m = droplet mass

V = droplet velocity

r = droplet radius

t = time

U = viscosity of the water

y = distance perpendicular to the surface

y = surface tension of the water

ea ' advancing contact angle and

Or = receding contact angle

The four terms in equation (3) account for the total force on the

droplet, viscous drag, surface-tension effects, and weight, respectively.

In order to calculate the maximum stationary droplet size, the

acceleration (or total force), and viscous drag terms can be set to zero

because under these conditions V = 0, and the resultant equation is

solved for r.

ir Y(cose -cosO ) +-2wr3 p 0 (4)crit a r 3 crit PS

1

3(cosO - Cosa) (2[3tC°Sr °Sa) 0.221 cm (5)

rcrit l 2pg

where rcrit = maximum stationary droplet radius (surface tension forces

balance weight), and p = liquid density.3

This critical size results in a droplet volume of 0.023 cm , and a

weight of 0.023 g. This size appears reasonable because a medicine

dropper typically produces droplets with a volume of 0.05 cm3.

-30-

If it is now assumed that a droplet that is 1% larger than the

critical size calculated above will flow down the rod, then the dynamic

force balance (Equation 3) can be solved for the flow velocity as a

function of position. First, a linear velocity gradient existing across

the droplet can be assumed by realizing the maximum velocity (Vm)

occurs at the outer edge of the droplet. Therefore,

dY Vmaxdy - r (6)

Use of this assumption effectively forces the entire droplet to travel

with velocity, Vmax. Integrating Equation 6 into Equation 3 results in:

2 2 dV 2 23r dt ~Y(eos6a- coser) 3 gp mz (7)

or

V

dt 1 2 m

where

3y(cosea - coser)C a - &2 +

2r2p

and

2 a2r p

(8)

-31-

The solution to this linear ordinary differential equation with the

initial condition that at t = 0, V = 0, has the form:

CI -C t-C (1-e ). (9)C2

Since V = dx/dt, integration of equation (9) yields the solution to

the distance traveled as a function of time

CI CI -C~tx 1 t + e 2 + K (10)

C2 C22

At t - 0, x 0 and

-Cl

C 22

For the conditions expected at Yucca Mountain,

C a 19.328 cmu/s 21C2 = 0.3012 /s

K = -213.05 cm.

Therefore,

x = 64.17t + 213.05e-0 .301 2 - 213.05 (11)

where x is in cm and t in seconds.

For vertical emplacement, the maximum height of uranium dioxide pellets

is 366 cm.

-32-

In order to apply these results to a horizontally emplaced spent-fuel

package and still maintain the actual geometric surface area of the rods,

the spent fuel was assumed to be a collection of 48 individual

0.82-cm-thick flat plates with a height of 42.5 cm. This height is

equivalent to the mean segment length (1.7R) of a circular cross section

with a diameter of 50 cm. Therefore, for horizontal emplacement, the

maximum height of the uranium dioxide is 42.5 cm.

The droplet retention times shown in the text in Table 2 were

iteratively calculated using these maximum distances and equation (11).

It is important to note that the surface coverages, which are based

on the contact times and-droplet-size, are very small. This result

implies that the inflowing water does not have to be uniformly

distributed across all spent fuel rods in each package and that the

spent-fuel pellets do not all have to be exposed.

References

Adamson, A. W., Physical Chemistry of Surfaces, Interscience Publishers,New York, NY, 1960, pp. 359-362.

Levich, V. G., Physiochemical Hydrodynamics, Prentice-Hall, Inc.,

Englewood Cliffs, NJ, 1962, pp. 669-672.

Sohn, H. Y., and M. E. Wadsworth, Rate Processes of Extractive Hetallurgy,Plenum Press, New York, NY, 1979, pp. 210-220.

Szekely, J., and M. J. Themelis, Rate Phenomena in Process Metallurgy,Wiley-Interscience, Mew York, fiY, 1971, pp. 57-62.

-33-

DISTRIBUTION LIST

B. C. Rusche (RW-l)DirectorOffice of Civilian Radioactive

Waste ManagementU.S. Department of EnergyForrestal BuildingWashington, DC 20585

Ralph Stein (RW-23)Office of Geologic RepositoriesU.S. Department of EnergyForrestal BuildingWashington, DC 20585

T. P. Longo (RW-25)Program Management DivisionOffice of Geologic RepositoriesU.S. Department of EnergyForrestal BuildingWashington, DC 20585

Cy Klingsberg (RW-24)Geosciences and Technology DivisionOffice of Geologic RepositoriesU. S. Department of EnergyForrestal BuildingWashington, DC 20585

J. J. Fiore, (RW-22)Program Management DivisionOffice of Geologic RepositoriesU.S. Department of EnergyForrestal BuildingWashington, DC 20585

B. G. Gale (RW-25)Siting DivisionOffice of Geologic RepositoriesU.S. Department of EnergyForrestal BuildingWashington, D.C. 20585

M. W. Frei (RW-23)Engineering & Licensing DivisionOffice of Geologic RepositoriesU.S. Department of EnergyForrestal BuildingWashington, DC 20585

E. S. Burton (RW-25)Siting DivisionOffice of Geologic RepositoriesU.S. Department of EnergyForrestal BuildingWashington, D.C. 20585

C. R. Cooley (RW-24)Geosciences & Technology DivisionOffice of Geologic RepositoriesU.S. Department of EnergyForrestal BuildingWashington, DC 20585

V. J. Cassella (RW-22)Office of Geologic RepositoriesU.S. Department of EnergyForrestal BuildingWashington, DC 20585

R. J. Blaney (RW-22)Program Management DivisionOffice of Geologic RepositoriesU.S. Department of EnergyForrestal BuildingWashington, DC 20585

R. W. Gale (RW-40)Office of Policy, Integration, andOutreach

U.S. Department of EnergyForrestal BuildingWashington, D.C. 20585

J. E. Shaheen (RW-44)Outreach ProgramsOffice of Policy, Integration andOutreach

U.S. Department of EnergyForrestal BuildingWashington, DC 20585

J. 0. Neff, ManagerSalt Repository Project OfficeU.S. Department of Energy505 King AvenueColumbus, OH 43201

-34-

D. C. Newton (RW-23)Engineering & Licensing DivisionOffice of Geologic RepositoriesU.S. Department of EnergyForrestal BuildingWashington, DC 20585

0. L. Olson, ManagerBasalt Waste Isolation Project OfficeU.S. Department of EnergyRichland Operations OfficePost Office Box 550Richland, WA 99352

D. L. Vieth, Director (4)Waste Management Project OfficeU.S. Department of EnergyPost Office Box 14100Las Vegas, NV 89114

D. F. Miller, DirectorOffice of Public AffairsU.S. Department of EnergyPost Office Box 14100Las Vegas, NV 89114

P. H. Bodin (12)Office of Public AffairsU.S. Department of EnergyPost Office Box 14100Las Vegas, NV 89114

B. W. Church, Director1{ealth Physics DivisionU.S. Department of EnergyPost Office Box 14100Las Vegasj XV 89114

Chief, Repository Projects BranchDivision of Waste ManagementU.S. Nuclear Regulatory CommissionWashington, D.C. 20555

Document Control CenterDivision of Waste ManagementU.S. Nuclear Regulatory CommissionWashington, D.C. 20555

S. A. Mann, ManagerCrystalline Rock Project OfficeU.S. Department of Energy9800 South Cass AvenueArgonne, IL 60439

K. Street, Jr.Lawrence Livermore NationalLaboratory

Post Office Box 808Mail Stop L-209Livermore, CA 94550

L. D. Ramspott (3)Technical Project Officer for MNWSILawrence Livermore NationalLaboratory

P.O. Box 808Mail Stop L-204Livermore, CA 94550

W. J. Purcell (RW-20)Office of Geologic RepositoriesU.S. Department of EnergyForrestal BuildingWashington, DC 20585

D. T. Oakley (4)Technical Project Officer for NNWSILos Alamos National LaboratoryP.O. Box 1663Mail Stop F-619Los Alamos, NH 87545

W. W. Dudley, Jr. (3)Technical Project Officer for NNWSIU.S. Geological SurveyPost Office Box 25046418 Federal CenterDenver, CO 80225

NTS Section LeaderRepository Project BranchDivision of Waste ManagementU.S. Nuclear Regulatory CommissionWashington, D.C. 20555

V. M. GlanzmanU.S. Geological SurveyPost Office Box 25046913 Federal CenterDenver, CO 80225

-35-

P. T. PrestholtNRC Site Representative1050 East Flamingo RoadSuite 319Las Vegas, NV 89109

M. E. SpaethTechnical Project Officer for NNWSIScience Applications

International CorporationSuite 407101 Convention Center DriveLas Vegas, NV 89109

SAIC-T&MSS Library (2)Science Applications

International CorporationSuite 407101 Convention Center DriveLas Vegas, UV 89109

W. S. Twenhofel, ConsultantScience Applications

International Corp.820 Estes StreetLakewood, CO 89215

A. E. GurrolaGeneral ManagerEnergy Support DivisionHolmes & Narver, Inc.Post Office Box 14340Las Vegas, NV 89114

J. A. Cross, ManagerLas Vegas BranchFenix & Scisson, Inc.Post Office Box 15408Las Vegas, NV 89114

Neal Duncan (RW-44)Office of Policy, Integration, and

OutreachU.S. Department of EnergyForrestal BuildingWashington, DC 20585

J. S. WrightTechnical Project Officer for NNWSIWestinghouse Electric CorporationWaste Technology Services DivisionNevada OperationsPost Office Box 708Mail Stop 703Mercury, NV 89023

ONWI LibraryBattelle Columbus LaboratoryOffice of Nuclear Waste Isolation505 King AvenueColumbus, OH 43201

W. H. Hewitt, Program ManagerRoy F. Weston, Inc.2301 Research Blvd., 3rd FloorRockville, MD 20850

H. D. CunninghamGeneral ManagerReynolds Electrical &

Engineering Co., Inc.Post Office Box 14400Hail Stop 555Las Vegas, NV 89114

T. Hay, Executive AssistantOffice of the GovernorState of NevadaCapitol ComplexCarson City, NV 89710

R. R. Loux, Jr., Director (3)Nevada Agency for Nuclear ProjectsNuclear Waste Project OfficeState of NevadaCapitol ComplexCarson City, NV 89710

C. H. Johnson, TechnicalProgram ManagerNevada Agency for Nuclear ProjectsNuclear Waste Project OfficeState of NevadaCapitol ComplexCarson City, NV 89710

-36-

John FordhamDesert Research InstituteWater Resources CenterPost Office Box 60220Reno, NV 89506

Department of ComprehensivePlanning

Clark County225 Bridger Avenue, 7th FloorLas Vegas, NV 89155

Lincoln County CommissionLincoln CountyPost Office Box 90Pioche, NV 89043

Community Planning andDevelopment

City of North Las VegasPost Office Box 4086North Las Vegas, NV 89030

City ManagerCity of HendersonHenderson, NV 89015

N. A. NormanProject ManagerBechtel National Inc.P. 0. Box 3965San Francisco, CA 94119

Flo ButlerLos Alamos Technical Associates1650 Trinity DriveLos Alamos, New Mexico 87544

Timothy G. BarbourScience Applications

International Corporation1626 Cole Boulevard, Suite 270Golden, CO 80401

Dr. Martin MifflinDesert Research InstituteWater Resources CenterSuite 12505 Chandler AvenueLas Vegas, NV 89120

Planning DepartmentNye CountyPost Office Box 153Tonopah, NV 89049

Economic DevelopmentDepartment

City of Las Vegas400 East Stewart AvenueLas Vegas, NV 89101

Director of CommunityPlanning

City of Boulder CityPost Office Box 367Boulder City, NV 89005

Commission of theEuropean Communities

200 Rue de la LoiB-1049 BrusselsBELGIUM

Technical Information CenterRoy F. Weston, Inc.2301 Research Boulevard,

Third FloorRockville, MD 20850

R. HarigParsons Brinkerhoff Quade &Douglas, Inc.

1625 Van Ness Ave.San Francisco, CA 94109-3678

Dr. Madan M. Singh, PresidentEngineers International, Inc.98 East Naperville RoadWestmont, IL 60559-1595

-37-

H. A. RevelliLawrence Livermore NationalLaboratory

P.O. Box 808Mail Stop L-206Livermore, CA 94550

E. P. BinnallField Systems Group LeaderBuilding 50B/4235Lawrence Berkeley LaboratoryBerkeley, CA 94720

V. H. OversbyLawrence Livermore NationalLaboratory

P.O. Box 808Mail Stop L-206Livermore, CA 94550

Jerry F. KerriskLos Alamos National LaboratoryMail Stop G787WX-4Los Alamos, NH 87545

6300 R. W. Lynch6310 T. 0. Hunter6310 CF 73-12142-1.61QIII6311 L. W. Scully6311 L. Perrine6312 F. W. Bingham6312 G. E. Barr6312 J. W. Braithwaite (10)6312 A. L. Dudley6312 H. S. Tierney6313 T. E. Blejwas6314 J. R. Tillerson6315 S. Sinnock6332 WHT Library (20)6430 N. R. Ortiz3141 S. A. Landenberger (5)3151 W. L. Garner (3)8024 P. W. Dean3154-3 C. H. Dalin (28)

for DOES/OSTI

-38-