ORNL-TM-13552.txt

 

MRV LT
1o s

ORNL/TM-13552

 

Strategy for the Future Use and
Disposition of Uranium-233:
Technical Information

P. J. Bereolos*
C. W. Forsberg
D. C. Kocher
A. M. Krichinsky

 

 

*Advanced Integrated Management Services, Inc.
 

This report has been reproduced directly from the best availeble copy.

Available to DOE and DOE contractors from the Office of Scientific and Techni-
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Available to the public from the National Technical Information Service, U.S.
Department of Commerce, 5285 Port Royal Rd., Springfield, VA 22161,

 

 

 

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ORNL/TM-13552

STRATEGY FOR THE FUTURE USE AND
DISPOSITION OF URANIUM-233:
TECHNICAL INFORMATION

P. J. Bereolos

ADVANCED INTEGRATED MANAGEMENT SERVICES, INC.
575 Oak Ridge Tumpike, Suite B-3
Oak Rudge, TN 37831

C. W. Forsberg
D. C. Kocher
A. M. Krichinsky

OAK RIDGE NATIONAL LABORATORY"
Oak Ridge, Tennessee 37831

April 1998

 

" Managed by Lockheed Martin Energy Research Corp., under contract DE-ACO05-
960OR22464 for the U.S. Department of Energy.
CONTENTS

LIS T OF FIGURE S oo v
LIS T OF TABLES oo v
ACRONYMS AND ABBREVIATION S oo vii
CHEMIC AL ELEMEN TS oo ix
PREF ACE e X1
EXECUTIVE SUMMARY ..o oo xiii
1. METHODS OF “°U PRODUCTION OR FORMATION ..o 1
1.1 NEUTRON BOMBARDMENT OF THORIUM .o 1

1.2 CONTAMINATION LEVELS OF 22U IN 23U oo 1

1.3 RADICACTIVE DECAY OF NEPTUNIUM ..o 2

2 DECAY CHAIN S oo 3
3. CHARACTERISTICS ..o VO TEN TR UTR TR TTTTRTT 7
3.1 CHEMICAL AND PHYSICAL PROPERTIES ..o oo 7
3 L Uranmum Meetal. 7

3 2 Urantum OXIAeS oo 7

3. 3 Uranmum FlOOmideS oo e 8

314 Uranyl NITTate . ....oooiiii i e 8

3.2 RADIOLOGICAL PROPERTIES . o oo e e 8
3.2.1 Comparison of “*Uand U........ooouooeoooeeeoeeeooee 8

3.2.2 Comparison with Other ISOtOPes............c..ooovveeeieiie e 10

3.3 BIOLOGICAL EEFEC T S oo e 14
3.4 NUCLEAR CRITICALITY oo 17

4. STORAGE REQUIREMENT S ..o oo oo 21
4. MATERIAL FORM oo 21
B2 PACKAGING ..o e 22
4.3 CONFINEMEN T 22
4.4 CRITICALITY CONT RO o oo 23
A5 SHIELDING oo e 24
4.6 SAFEGUARD S oo 26
4.6.1 DOE ReqUITements ...........oooooiiii e 26

4.6.2 JAEA ReqUIrements ... 27

4.6.3 Elmination of Weapons Potential ............................................... 29

5. DISPOSAL REQUIREMENTS FOR *’U DECLARED AS WASTE ... 31
5.1 RADIOACTIVE WASTE CLASSIFICATION ..o 31
5.2 CLASSIFICATION AND DISPOSAL AS HAZARDOUS WASTE ... 32

111
5.3 RADIOACTIVE WASTE DISPOSAL OPTIONS ... 33
5.3.1 Near-Surface Disposal of LLW ... 33

5.3.2 Alternatives to Near-Surface Disposal ... 34

5.4 MIXING OF U WITH OTHER WASTES ... 35
5.5 NUCLEAR CRITICALITY ..o 36
5.5.1 Need to Avoid Criticality.................ocoovieeiii 36

5.5.2 Nuclear Cnticality Control by Isotopic Dilution............................... 37

6. REFE RENCES . e 39

v
LIST OF FIGURES

Fig. ES.1. National strategy for future use and disposition of “*U. ... XV
Fig. 2.1. Decay chain 0f 22U . oo, 4
Fig. 2.2. Decay chain of Z2U. ... 5

Fig. 3.1. Gamma exposure at 1 ft from 10 kg of UO; with varying amounts of *?U......... 9

Fig. 3.2. Alpha activity and gamma exposure rate at 1 ft as a function of time calculated
for 1 kg *U (with 100 ppm *?U) as a loose-pour powder (1.5 g/cm’) contained in a

3-in. diam. by 6-in. tall can with 20-mal-thick steel walls......................... 11
Fig. 4.1. Lead shielding for U with various concentrations of ®?U. ................c..cc...... 25
LIST OF TABLES

Table 2.1. Half-lives, branching fractions, and principal decay modes for isotopes of
uranium and for “*Th and their short-lived decay products ...............ccocococeoovenrnnnn. 6

Table 3.1. Selected radiological data for isotopes of uranium and thorium and their

short-lived decay products and for “Puand *'Am........._._.................. 12
Table 3.2. Limits on activity concentrations in air and water for releases to the

environment for isotopes of uranium and thorium and for “’Pu and *'Am............ 15
Table 3.3. Values of basic nuclear safety parameters.............ooooovniii e, 18
Table 4.1. DOE nuclear material safeguards categories .........coooo.ocvviviiiiinniiiec e, 26
AEA
ANS
ANSI

CFR
C/S

DIQ
dis
DOE
pU

EIS
EPA

GCD

HEU
HLW

IAEA
ICRP

LEU
LLW

NRC
NTS
NWPA

ORNL
RCRA
SCALE
S1

SNF
SRS

TIDs
TRU

WAC
WIPP

ACRONYMS AND ABBREVIATIONS

Atomic Energy Act
American Nuclear Society
American National Standards Institute

Code of Federal Regulations
containment and surveillance

Design Information Questionnaire
disintegrations

U.S. Department of Energy
depleted uranium

Environmental Impact Statement
U.S. Environmental Protection Agency

Greater Confinement Disposal

highty enriched uranium
high-level waste

International Atomic Energy Agency
International Commission on Radiological Protection

lowly enriched uranium
low-level waste

U.S. Nuclear Regulatory Commission

Nevada Test Site

Nuclear Waste Policy Act

Oak Ridge National Laboratory

Resource Conservation and Recovery Act

Standardized Computer Analyses for Licensing Evaluation
International System of Units

spent nuclear fuel
Savannah River Site

tamper indication devices
transuranic

waste acceptance criteria
Waste Isolation Pilot Plant

vii
CHEMICAL ELEMENTS

Actinium
Americium
Astatine
Beryllium
Bismuth
Fluorine
Francium
Hydrogen
Nitrogen
Neptunium
Oxygen
Silicon
Protactinium
Lead
Polonium
Phitomum
Radium
Radon
Thorium
Thallium
Uranium

ix
PREFACE

This report is one 1n a serigs of reports which examines issues associated with the
future use and disposition of “?U. A brief description of the other reports is included
herein.

ORNL/TM-13550, Strategy for the Future Use and Disposition of Uranium-233:
Overview. This document is a summary of the path forward for disposition of excess
“3U. It includes required activities and identifies where major programmatic decisions
will be required.

ORNL/TM-13551, Strategy for the Future Use and Disposition of Uranium-233:
History, Inventories, Storage Facilities, and Potential Future Uses. This document
includes the sources, historical uses, potential future uses, and the current inventory of
*3U. The inventory includes quantities, storage forms, and packaging of the material.

ORNL/TM-13553, Strategy for the Future Use and Disposition of Uranium-233:
Options. This document describes the proposed disposition alternatives, the technical
advantages and disadvantages of each option, and the institutional issues associated with
each option.

ORNL/TM-13524, Isotopic Dilution Requirements for *>>U Criticality Safety in
Processing and Disposal. This document analyzes criticality issues associated with
processing and disposing of “U.

ORNL/TM-13517, Definition of Weapons-usable Uranium-233. This document
develops a definition of non-weapons-usable “*U to provide a technical basis for
changing the safeguards and security requirements for storing, using, and disposing of
U that is isotopically diluted with Z*U.

X1
EXECUTIVE SUMMARY

- This document provides a summary of technical information on the synthetic
radioisotope “U. It is one of a series of four reports that map out a national strategy for
the future use and disposition of U (Fig. ES.1). The technical information on “*U in
this document falls into two main areas. First, material characteristics are presented
along with the contrasts of “*U to the more well known strategic fissile materials, Z°U
and plutonium (Pu). Second, information derived from the scientific information, such as
safeguards, waste classifications, material form, and packaging, is presented.
Throughout, the effects of 1sotopically dllutmg ‘3’U with nonfissile, depleted uranium
(DU) are examined.

Minute amounts of >°U are formed as a decay product of *’Np in spent nuclear fuel
(SNF). However, the matenal under consideration in this report has been intentionally
produced by the bombardment of natural thorium with neutrons in nuclear reactors.
Uranium-233 is a long-lived isotope with a half-life of 1.592 x 10° vears. It decays
directly to **Th, which is also relatively long-lived with a half-life of 7340 years. For the
same mass of material, the alpha activity of “*U is more than three orders of magnitude
greater than that of U and about one order of magnitude less than that of **Pu.

A significant factor in the production of U from thorium is the formation of Z*U,
which is an undesirable contaminant isotope. The presence of even small amounts of “°U
is important in determuning the radlologlcal properties of materials consisting mainly of
uranium. The decay chain of U is quite different from that of “°U. Although » 2U is
the longest-lived 1sotope 1n its decay chain, 1t has a half-life of only 68.9 years.
Therefore, in planning for disposal of Z°U, the amount of 2U contamination becomes
insignificant. However, for interim storage, handling, and use, the decay chamn of B2U
presents several complications. The primary consideration is the decay product 2T,
which emits a 2.6-MeV gamma ray. The quantities of ***U present with “°U determine
the radiation shielding requirements, and sigmficant shielding is usually needed. Also
included in the decay chain of “*U is ’Rn, which exists as a gas under standard
condifions. Therefore, storage facilities must have adequate delay times in ventilation
systems to prevent this matenal from escaping before 1t has decayed to a filterable
particulate form. Finally, ®*U lacks the equivalent of a long-lived “stopper” isotope like
?2Th that can be used to “break” the decay chain through a chemical separation. ' Only
very brief periods (1.¢., weeks) of relief from penetrating gamma emitters can be realized
by removing “*Th (1.9-year half-life), the first decay product of “*U.

Uranium and its compounds can cause biological damage both chemically and
radiologically. It is in the radiological properties of ~°U that one sees important contrasts
with other isotopes of uramium. Uranium-233 has a higher specific activity than does
35U or natural uranium. Additionally, **U almost always contains “**U, with its much
shorter half-life, very high specific activity, and associated gamma emissions. Therefore,
radiation damage to humans exposed to a given mass of *~U is potentially much more
severe, relative to exposure to the same mass of the other prevalent uranium isotopes.
Even in the absence of *?U, one g of *’U has the same radiological significance as

xiii
15¢ 24U, 4400 g U, 150 g ¢ or 28000 g P*U. In all materials of concem to this
report, ~~U is the most important long-lived radionuclide.

Uranium can form a variety of chemical compounds. Triuranium octaoxide (U;Os) 1s
the most thermodynamically stable form in dry air and is the preferred storage form.
Uranium 1s also stored in a variety of other chemical forms including uranium metal,
oxides other than U;QOs, and fluondes. Uranium-233 and other uranium 1sotopes have no
significant differences in their chemical and physical properties—except for the effects of
greater levels of radiation on chemical compounds associated with the “*U.

Because 2°U is readily fissionable, nuclear criticality is also an important concern
during storage and disposal. The minimum critical mass of **U is less than that of **°U.
While 1n storage, criticality may be controlled primarily through geometry. However,
during long-term disposal, geometry cannot be guaranteed. Therefore, isotopic dilution
with DU (99.8 wt % U and 0.2 wt % *U) becomes an attractive altemative for
criticality control. Dilution of “°U to ~0.53 wt % with DU (a) minimizes the potential for
long-term criticality, (b) is equivalent to ~1 wt % “°U, and (c) is the criticality limit for a
2 system in a homogeneous mixture containing water, uranium, and silicon oxide.

Isotopic dilution also helps prevent nuclear proliferation. Because U is fissile, it
may be used to produce nuclear weapons. Isotopic dilution of “°U to 20 wt % is already
the preferred method for 2’U demilitarization in the United States. Technical analysis
indicates that isotopic dilution to ~12 wt % *°U (a) minimizes the potential for use of
U in nuclear weapons and (b) is equivalent in terms of nuclear weapons use to 20 wt %
U, However, unlike the situation with U, there is neither federal regulation nor
international agreement on the amount of isotopic dilution necessary for ***U to minimize
its weapons potential. Therefore, all “’U needs continual safeguards — physical
protection, surveillance, and accounting.

If **U were declared a waste, it would be classified presently as low-level waste
(LLW). However, waste containing significant amounts of ***U probably would not be
suitable for shallow-land disposal. The disposal options for such waste include (after
1sotopic dilution with DU) a geologic repository or a greater confinement disposal facility.
The latter type of facility has been operated in the past at the Nevada Test Site.
Processing of ’U with other wastes presents geologic disposal options. If the **U were
processed with high-level waste (HEL W), the resulting waste would be classified as HLW.
If the *U were processed with transuranic (TRU) waste, the resulting waste would be
classified as TRU waste provided that the concentrations of long-lived alpha-emitting
transuranic radionuclides still exceeded 100 nCi/g. Wastes containing *>>U might also be
regulated as solid hazardous waste under the authority of the Resource Conservation and
Recovery Act (RCRA). Indeed, it is a U.S. Department of Encrgy (DOE) policy to
manage all of its radioactive waste according to the requirements of RCRA, unless such
waste 1s found not to be hazardous, as defined under RCRA, depending on its non-
radiological properties.

X1V
 

U-233 from Qutside Facilities

 

 

Inventory ‘L Consolidation

 

Existing Short-Term -y
U-233 Storage

 

 

Programmatic Uses and
| Material Treatment Needs

ORNL DWG 98-3893

 

 

Assessment of

 

 

) Dilution/Treatment/Packaging

Processes

 

Off-Gas Treatment

  

 

 

 

Short- or Long-Term
Storage and Use

 

I¢

 

 

 

 

U-233 with potential future uses

Long-Term Storage or Disposal
in a Geological Repository

 

Surplus U-233

 

 

 

Fig. ES.1. National strategy for future use and disposition of *°U.

Xv
1. METHODS OF **U PRODUCTION OR FORMATION

Uranium-233 1s a synthetic isotope discovered in the early 1940°s by John Gofman at the
University of California, Berkeley (Gofiman 1943). It is a fissile isotope that can be used in
nuclear reactors to generate heat and electricity. In isotopically purer concentrations, it can be
used in nuclear weapons. Uranium-233 is produced by neutron bombardment of natural thorium,
and it also is the decay product of long-lived *’Np. During neutron bombardment of thorium, =*U
is also formed in various concentrations, and its presence usually governs precautions that must be
taken while handling the main product, **U.

1.1 NEUTRON BOMBARDMENT OF THORIUM

The principal method of producing ***U is by bombarding *Th with neutrons in an
accelerator or a reactor:

232Th .Y ; 233Th [3-" 233Ppg fiw » 233)J

22.3 min 27.0d

The Thorex solvent extraction process is used to separate uranium from spent thorium fuel.
Typically, the Thorex process removes more than 99 % of the “>Th in the spent fuel. Therefore,
**Th is usually present only in small quantities in ~°U materials.

1.2 CONTAMINATION LEVELS OF Z*U IN **U

Uranium-232 is another synthetic isotope of uranium formed along with “’U during
irradiation of ““Th and “Th within a reactor. Although “Th is not found in significant mass
abundance in nature, its concentration in thorium fuel influences the lower bound on “*U formation
in reactors. The formation of “*U in thorium fuel is shown below:

n,2n
Y, n - -
252 — Y #'Th -—-—-»B 231 ——»n, ! #2Pa -——-)-B 232
Th | 1.0633 d Pa 131d u
n,vy T n,2n T
230Th 233U

Both the amount of **U and the ratio of **U to *°U produced increase with increasing
neutron flux and irradiation time. The minimum energy threshold for the (n,2n) and (v,n) reactions
is 6.34 MeV, so ~*U formation is largely dependent on the neutron and gamma energy distribution
in the reactor (Till 1976; Meichle 1963).
With increasing knowledge about thorium, several methods were developed to produce low-
contaminant “°U at a reasonable cost. A simple improvement was to avoid using ores that are rich
in 2°Th. This avoidance reduced one reaction pathway for producing **U. Ores with low *°Th
concentration are readily available (c.g., monazite). Another way of reducing *U production was
to lower the exposure of the thorium targets to high-energy neutrons. There were two ways to
accomplish this task. First, the reactor was loaded so that the targets were exposed to only a low-
energy neutron flux. Second, the use of short irradiation times minimized the fissioning of newly
formed “*U and the consequent production of high-energy fission neutrons and gamma rays in
close proximity to the still-fertile natural thorium. Short irradiation times also reduced the heat
generation in the target, thus allowing methods to cheapen the target fabrication process, such as
using thorium oxide instead of thortum metal (Boswell et al. 1966).

For a single-core fueling cycle under reactor conditions, the resultant ***U concentration would
be typically less than 0.05% by mass of total uranium. Multiple cycles could build the U
concentration up to 0.15%. Under low-power, weapons-production reactor conditions, ~*°U
concentrations were held to as low as 5 ppm (on a total uranium basis) for an irradiation cvcle.

Although *?U has a slight tendency to fission upon neutron capture, its dilute concentration
(and its association with highly fissile “*U) presents an insignificant contribution to nuclear
criticality.

1.3 RADIOACTIVE DECAY OF NEPTUNIUM

Small amounts of **U are produced by the decay of “’Np. Neptunium-237 is a decay
product of *’U, which is produced in nuclear reactors primarily through multiple neutron capture
by “°U and, to a lesser extent, (n, 2n) reactions with *U. Therefore, *°U will be present in spent
nuclear fuel (SNF). However, because *’Np is long-lived (2.2 x 10°-year half-life), only a small
amount of *’U will be produced in this manner before disposal of the material. Because separating
the *U from the other uranium isotopes present in reactor fuels is very difficult, SNF is generally
not considered as a source for “°U.
2. DECAY CHAINS

The decay chain of “*U is a part of the Neptunium Series. Uranium-233 is a long-lived
isotope (1.592 x 10°-year half-life) that decays to a series of alpha-emitting and beta-emitting
radionuclides (Fig. 2.1). lts first decay product, “*Th, also has a long half-life (7340 years).
These long half-lives mean that the decay products after “*Th will not be present in significant
quantities during short-term storage and handling. The remaining decay products in the chain are
relatively short-lived. Three isotopes (*'Fr, “’Bi and **TI) in this series also emit significant
intensities of higher-energy gamma rays. The decay chain ends with the stable product **Bi.

One of the most significant characteristics of ““U is its decay chain (Fig. 2.2). Uranium-232
has a short half-life of 68.9 years followed by the even shorter half-hived series of mostly alpha-
emitting decay products. Because of the short half-life of U, its decay products are present soon
after production. The last member of this decay chain, “*T1, emits a beta particle accompanied by
a highly energetic (i.e., extremely penctrating) gamma emission (2.6 MeV). Other, less energetic
gamma emissions from *'’Bi are also of concern, although they occur at considerably lower yields
than does the ***T1 emission. The presence of ““U and the gamma emissions associated with its
decay chain dictate many of the precautions required in handling ?*U.

Another hazard associated with the “*U decay chain is the presence of “’Run. At normal
temperatures and pressures, radon exists as a gas. As a gas, “’Rn can cause problems during both
storage and handling because of its mobility.

Table 2.1 (based on Browne, Firestone, and Shirley 1986) lists the half-life of each long-lived
isotope of uranium and its shorter-lived radioactive decay products, the branching fraction for each
short-lived decay product in the decay of its long-lived parent isotope, and the principal decay
modes for each isotope and its decay products. These data are also given for “*Th. For purposes
of comparison, data for “*Pu and **' Am, which are important alpha-emitting transuranic (TRU)
1isotopes in high level waste (HLW) and TRU waste, are also included.

The shorter-lived decay products listed with each isotope of uranium or thorium in Table 2.1
are those that would achieve secular equilibrium with the long-lived parent isotope within a short
period of time after chemical separation. Therefore, these decay products generally would be of
concern in determining the radiological properties of any materials containing uranium or thorium
during handling or disposal. The branching fraction for each decay product determines its activity
relative to the activity of the long-lived parent isotope at equilibrium.

Except for 22U, the decay chain for each isotope of uranium also includes a long-lived
“stopper” radionuclide that can be used to “break™ the decay chain by a chemical separation.
Specifically, U decays to “*Th(T,» = 7340 y), ®*U to “*Th(T;, = 7.54 x 10*y) , U to
BIpa(Tip=3.276x10% y), “°Uto **Th(T1» = 1.405 x 10'° v), and =*U to *U(T ., = 2.454 x 10°
v). Because of their long half-lifes, these decay products will not be important in determining
radiological properties during handling of chemically separated uranium. Except for the presence
of *Ra and **Ac, the decay chain for “*Th is the same as is that for “?U. For comparison, the
isotopes, 2Pu and **' Am, decay to the longer-lived “stopper” isotopes, °U and “"Np,
respectively. Due to their long half-lives, these decay products are not radiologically significant.
o 2.16%

U-233
1.592x 10 y
a
Th-229 Ac-225
7340y 10.0d
o o
B-
Ra-225 Fr-221
14.8d 49m
a
At-217
0.0323 s
o
Bi1-213
4559 m
T1-209
220m

ORNL DWG 98-3394

Gamma Emissions of Note

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Yield
Nuclide Energy { keV) {emissions/dis)
U-233 317.133 0.0088
320.508 0.0031
Th-229 193.59 0.046
21094 0.033
Ra-225 40.34 0.29
Fr-221 21798 0.109
Bi-213 44034 0.165
Ti-209 117. 0.81
467. 0.81
1566. 0.98
Po-213 Bi-209
42 us stable
x
Pb-209
3253

Fig. 2.1. Decay chain of *’U.

 
ORNL DWG 98-3895

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

U-232 Gamma Emissions of
689y
Yield
o Nuclide Energy (keV) | (emissions/ dis) |
Pb-212 238.578 0436
300.034 0.0334
Bi-212 72725 0.0665
Th-228
1.913 y 785.51 0.01107
1078.69 0.00535
o 1620.66 0.0151
T1-208 27728 0.068
510.606 0216
Ra-224 583.022 0.86
3.66d 763.06 0.0164
860.3C 0.120
o 2614.35 0.9979
1;‘;%20 Po-212
98 0.298us
a
o
B 64%
Po-216 Bi-212
0.150 s 1.0092 h Pb-208
stable
o o 36%
fi_
Pb-212 T1-208
10.64 h 3053 m

Fig. 2.2. Decay chain of %,
Table 2.1. Half-lives, branching fractions, and principal decay modes for isotopes of wranium

and for

Th and their short-lived decay products

 

 

S0P ez roduce marie® OO ey modes
2y 6.89 x 10'y Alpha
28Th 1913 x 10°y 1.0 Alpha
2'Ra 3.66 x 10°d 1.0 Alpha
Rn 556 x 10" s 1.0 Alpha
215pg 1.50 x 107" s 1.0 Alpha
22pp 1.064 x 10' h 1.0 Beta/gamma
224 1.0092 x 10°h 1.0 Alpha, Beta/gamma
22pg 2.98 x 107" ps 0.6407 Alpha
2571 3.053 x 10°m 0.3593 Beta/gamma
U 1.592 x 10°y Alpha
24 2.454 x 10°y Alpha
2y 7.037 x 10%y Alpha
BiTh 1.0633 x 10° d 1.0 Beta/gamma
2y 2342 %107y Alpha
2y 4468 x 10°y Alpha
Th 2410 % 10'd 1.0 Beta/gamma
PPy 1.17x 10°m 1.0 Beta/gamma
B4pa 6.70 x 10°h 0.0016  Beta/gamma
B2y 1.405 x 10"y Alpha
“Ra 575 % 10° v 1.0 Beta
Bac 6.13%10° h 1.0 Beta/gamma
ZZBThu‘
pu° 2411 x 10%y Alpha
I Am* 4327 x 10°y Alpha

 

2 Short-lived decay products are those with half-lives of a few vears or less which normally should be
present and in activity equilibrium with long-lived parent isotope shortly after chemical separation.
?Values from Browne, Firestone, and Shirley 1986.

‘ Number of atoms produced per decay of long-lived parent isotope.

4 Data for **Th and its decay products are listed following entry for ~*U.
“ Data for *°Pu and **' Am are provided for comparison only. These isotopes would not be present in
chemically separated uranium.
3. CHARACTERISTICS

3.1 CHEMICAIL AND PHYSICAL PROPERTIES

Uraninm exists as a pure metal, and because of its strongly electropositive nature, it forms
compounds with all nonmetallic elements except for the noble gases. Uranium has four oxidation
states in aqueous media: +3, +4, +5, and +6. The U state is very unstable with respect to
oxidation and is a red-wine color. U™ reduces water, yielding U™ and hydrogen. U™ (known as
the uranous 1on) 1s metastable with respect to oxidation by nitrate and is a degp-green color. The
+5 state, UQ,", tends to disproportionate to U™ and U0, The +6 state, UO,™ (uranyl ion), is
yvellow and is the most prevalent and important aqueous state. It can be reduced to the +4 state
chemically, photochemically, or electrochemically.

3.1.1 Uramum Metal

Pure uranium is a heavy metal that exists as silver-white or black crystals. It is ductile and
malleable (Uranium Storage Assessment Team 1996). It melts at 1132°C, boils at 3818°C, and
has a density of 19.04 g/cm’. (By comparison, lead melts at 327.3°C, boils at 1750°C, and has a
density of 11.35 g/cm’). When uranium metal is in the form of solid chips, shavings, or dust, it
can be a dangerous fire hazard if exposed to heat or flame in air. Uranium dust can also be an
explosion hazard if exposed to flame 1n the presence of oxygen.

Uranium metal can react vigorously, even violently, with oxidizing agents. Solid pieces,
larger than 1/16-in. diam., will not spontaneously ignite (Peacock 1992), but their surfaces will
corrode. The corrosion rate depends on surface area, temperature, humidity, and the presence or
absence of oxygen. Corrosion of uranium metal has two primary consequences. First, it converts
a cohesive metal sohd to a dispersible oxide dust. Also, under humid conditions, a by-product of
corrosion, hydrogen, can lead to a fire or an explosion hazard or can contribute to container
pressurization.

3.1.2 Uranium Oxides

Uranium oxides are the most significant compounds with regard to storage. The uranium-
oxvgen phase diagram is complex. Many binary oxides and crystalline modifications have been
reported. Three of the uranium oxides are common in ?’U processing and storage areas: uranium
dioxide (UOy), uranium trioxide (UOs), and triuramum octaoxide or pitchblende (U;0y), whichis
sometimes simply referred to as urantum oxide.

Urantum dioxide 1s the most common compound used (in a compressed pellet form) in reactor
fuels and is a significant intermediate in metal manufacture. It exists as brown-black or sometimes
green-black crystals that are fairly stable chemically. At high temperatures, nonstoichiometric
forms exist with vanable oxygen ratios ranging from UQ, g to UO, 5. In very finely divided form,
U0, 1s potentially pyrophoric. | |

Another significant intermediate in metal manufacture is UO;. It is a yellow-red powder that
is chemically stable, except for hydrate formation, and is routinely prepared by thermal
decomposition of nitrate or peroxide.
The most stable oxide 1s U;QOs, an olive-green powder. Its stability makes 1t best suited for
long-term storage (Cox 1993). It 1s the primary oxide formed by buming (above 650°C) 1n excess
atr and by corrosion after extended air exposure, so it can be derived readily from the other oxides.
Because UsOg has more uranium atoms per hydrogen atom than the other two prevalent uranium
oxides, proportionally less moderation i1s provided by waters of hydration.

3.1.3 Uranium Fluorides

Uranium fluorides are extensively used in the U fuel cycle to enrich natural uranium.
However, fluoride compounds have less significance for the synthetic **U. Uranium tetrafluoride
(UF.) is nonvolatile and was used in the Molten Salt Reactor Experiment. However, HF is often
chemically absorbed on UF,. This absorption can cause storage problems by accelerating
corrosion of storage packages. Also, UF, can be directly fluorinated to form uranum
hexafluoride (UF;), which is volatile. Uranmium hexafluoride is highly reactive with water and moist
air, forming uranyl fluoride (UO-F,) and releasing hydrogen fluoride, both of which are chemically
toxic. Inhalation and ingestion of UK, result in acutely serious health threats. Consequently, UFg
must be stored in gas-tight, corrosion-resistant canisters.

3.1.4 Uranyt Nitrate

Uranyl nitrate solution, UO2(NO3),, is an important intermediate in the purification of
uranium by solvent extraction. It is formed by the aqueous reaction of nitric acid (HNOs) and
uranium oxides. It forms a yellow cake that corrodes iron cans and degrades some plastics.
Uranyl nitrate solutions can be absorbed through the skin.

3.2 RADIOLOGICAL PROPERTIES

The radiological properties of any material depend on the activity of various radioisotopes
that arc present. The activity of a radioisotope is defined as the number of disintegrations (dis) per
unit time. The conventional unit of activity is the curie (Ci), which is defined as 3.7 x 10'° dis/s,
and the International System of Units (SI) unit of activity 1s the becquerel (Bq), which is defined as
1 dis/s. The activity of any radioisotope is related to its mass by its specific activity.

3.2.1 Comparison of >**U and **U

The most important factor that determines the external radiation field for “U is the quantity
of ?U present, because of the high-energy gamma radiation emitted by the **U decay product
“*TI. Figure 3.1 shows the calculated radiation levels over time (after chemical separation to
interrupt the decay chain producing **TI) at several concentrations of **U (Krichinsky 1975).
These calculations were made for a distance of 1 ft from 10 kg of UO; packed m a cylindncal can
with a 6-cm radius, 12-cm height, and 12-mul wall thickness. After the initial increase as the
activity of gamma-emitting decay products increases, the radiation levels are roughly lmearly
proportional to concentration of “*U. The maximum levels are reached after about 10 years.
Gamma expesure rates (R/hr)

ORNL DWG 98-3896

1000

 

x

 

 

 

 

 

 

 

 

 

 

— ®
._1__ &

0.01 - : : ', : z | : ;
0 200 400 600 800 1000 1200 1400 1600 1800

Time after purification (days)

Fig. 3.1, Gamma exposure at 1 ft from 10 kg of UO, with varying amounts of >y,

2060

 

X 1000 ppm U-232
@ 100 ppm U-232
A 10 ppm U-232

¢ lppmU-232 |

  
 
 
  

 
In addition to the gamma activity of the U decay chain, the gamma activity of *°U itself and
of residual fission products must be considered. Although “*U is principally an alpha emitter,
penetrating gammas are produced in **U decay, primarily in the 40- to 320-keV region. However,
because of their relatively low energies and intensities, these gamma emissions can easily be
shielded. For the residual fission product activity, a practical goal seems to be about 10° dis/min
per gram “°U that produce 0.5-1.0 MeV gamma rays. This is about the minimum activity to be
expected in the Thorex process product (Amold 1962). However, the residual fission product
content of “’U can be decreased to almost any desired level by decontamination beyond that
obtained 1n the Thorex process.

In the long term, in comparison with ***U, the quantity of **U becomes less of a factor
because of its short half-life. Figure 3.2 shows the long-time radioactivity for 1 kg of **U with
100 ppm of **U. Despite the relatively high initial concentration of *?U, after 50 years the
contribution to alpha activity from the ***U and U chains is roughly equivalent. After 500 years,
the radioactivity from **U and its decay products is negligible, while the radioactivity of the U
chain is still increasing. This 1s of major importance when considering disposal in a geologic
repository.

3.2.2 Comparison with Other Isotopes

In regard to the radioisotopes of greatest importance, materials containing high concentrations
of *°U are unusual compared with more familiar types of radicactive waste containing high
concentrations of alpha-emitting isotopes (1.e., HLW and TRU waste). Thus, it is useful to
compare radiological data for U and other isotopes of uranium that may be present with the data
for other alpha-emitting isotopes which commonly occur in radioactive wastes. In addition,
because the abundances of different isotopes are usually reported in terms of mass rather than
activity, it 1s useful to discuss the relationship between mass and activity abundances of different
isotopes of uranium in order, for example, to identify the mass abundance of ***U at which this
1sotope would present the greatest radiological concern in the materials.

The basic radiological data for **U and the other longer-lived isotopes of uranium which may
be present in materials containing > U are summarized in Table 3.1 [based on Schleien (1992),
Kocher (1980), and Unger and Trubey (1981)]. Data also are included for the alpha-emitting TRU
izgzotopes **Pu and **' Am, which are important long-lived isotopes in HLW and TRU waste and for
Th.

Table 3.1 lists the specific activity, thermal power, specific gamma-ray dose constant, and
mean gamma-ray attenuation coefficient in lead. These radiological data are described in the
following paragraphs.

The specific activity listed in Table 3.1 is defined as the activity per unit mass of the given
isotope. Thus, shorter-lived radionuclides have relatively high specific activities and longer-lived
radionuclides have lower specific activities. Only the specific activities of the long-lived parent
1sotopes of uranium or thorium are listed, because the activity of any shorter-lived decay product at
equilibrium is determined by the initial activity of its long-lived parent isotope and the branching
fraction for the decay product, as given in Table 2.1.

10
I

Gamma Exposure Rate at 1 ft (R/h)

ORNL DWG 98-3897

 

 

 

 

 

12 80000
G amma Exposure Rate at 1 ft (R/h) ™\ Alpha Activity
~== {J-232 chain alpha radioactivity f \
/ \ 1 70000
10" 2= U-233 chain alpha radioactivity l
—u@ e Total alpha radioactivity f \
Gamma Activity * 1 60000
8 + X
l \\ 4 50000
/ \
6+ / { + 40000
/
7
A + 30000
47 - /
Soom
A ~ " / \
, - - \  + 20000
& . B e, i
24 g > ~a, }J’" \
%, - »
W om om o= owmin s m wh b-*:fl*-’.*ll#l&:* \ 4+ 16000
M SBW Sfl\\ ’ x
] T P e " ,%
0 '. ! ! ! G e ko hamomn b B3 W ()
1.E-02 1.E-01 1.E+00 1.E+01 1.E+02 1.E+03 1.E+04 1.E+05 1.E+06
Time (y)

Fig, 3.2. Alpha activity and gamma exposure rate at 1 ft as a function of time calculated for 1 kg

loose-pour powder (1.5 glcms) contained in a 3-in. diam by 6-in, tall can with 20-mil-thick steel walls,

230 (with 100 ppm 211y as a

Alpha Radioactivity (Ci)
Table 3.1. Selected radiological data for isotopes of uranium and thorium and their short-lived
decay products and for **°Pu and *'Am

 

Specific activity

 

Thermal power r 1
Isotope (Cilg)® (W/g)* (ren/h-uCi)* (cmy
By 2.1 % 10 6.9 x 107 89 x10°® 7.2 x 102
(5.2)¢ (1.3 x 1076

28Th 7.1 x 107 7.9 x 107 7.3 x 10°

22pa 7.4 %x 107 1.1x 1078 8.1

22%Rn 8.2 x 107 3.6 x 1071° 1.5

26pg 8.8 x 107 9.0 x 10712 97 x 107

22py 4.2 % 1072 2.7 % 107 1.1 x 10"

2p; 3.7 x 107 1.9 x 107 1.0

2p, 73 % 107 0.0

20811 1.9 x 107 6.1 x 107 55x 107
3y 9.6 x 107 2.8 x 107 2.9 x 107® 7.1 x 10°
By 6.2 % 1073 1.8 x 107 78 x 1078 7.2 % 107
35U 2.2x107° 6.0 x 107 3.4 x 107 2.3 x 10!

(6.3 x 107%y? (8.8 x 107)?

BiTh 2.4 x107° 5.5x 107 6.8 x 10°
35y 6.5 x 107 1.8 x 107 7.4 x 107 7.2 x 10?
B3y 3.4 x 107 86 x 107 6.5x 107% 7.2 x 10°

(1.0 x 10757 (1.5 x 107?

Z4Th 1.4 % 107'° 7.5 x 1078 1.5x10°

B4mpy 1.7 x 107° 1.0 x 107% 9.8 x 10

Bipy 8.0 x 1072 3.2 x 107 1.1

(table continued on next page)

12
Table 3.1. Selected radiological data for isotopes of uranium and thorium and their short-lived
decay products and for ®°Pu and **'Am (continued)

 

 

Specific activity ‘Thermal power I M
Tsotope * (Ci/g)’ _ (wrgy (rem/h-pCi)* (cm?y
Z2Th 1.09 x 107 2.7 x 107 6.8 x 10°® 8.3 x 10?
(2.7 x 107%? 2.1 x 1078
*®Ra 7.7 x 10712 0.0
“ZBAc 9.0 x 107° 8.4 x 107 9.9 x 107
“5Th 3.6 x 107 7.9 x 107° 7.3 x 10°
*¥Ra 3.8x 107 1.1 x 107 8.1
Z'Rn 42x107° 3.6 x 10710 1.5
2o 4.5 x 107° 9.0 x 1072 9.7 x 107
*12pp 2.1x107° 2.7 x 1077 1.1 x 10!
22g4 1.9x 107 1.9 x 107 1.0
2pg 3.8x10°° 0.0
298] 95x 1077 6.1x 1077 55x 107
%Py 62 x 1072 2.0 x 107 3.0 x 107° 8.6 x 10°
Am 3.4 1.2 x 107 3.1 x 107 2.6 x 102

 

? Indented entries are short-lived decay products listed in Table 2.1, which are assumed to be in
activity equilibrium with long-lived parent isotope.

® Activity per unit mass of long-lived parent isotope obtained from Table 8.4.1 of Schieien (1992).
At equilibrium, activity of each short-lived decay product per unit mass of long-lived parent isotope is
equal to specific activity of parent multiplied by branching fraction for decay product given in Table 2.1.

“Values are given per unit mass of long-lived parent isotope and are based on total energy of all
ionizing radiations per decay of particular isotope given in Kocher (1980), energy of recoiling decay
product nucleus, specific activity of parent isotope, and branching fractions for short-lived decay products
given in Table 2.1,

“Value is total for long-lived parent isotope and its short-lived decay products when all decay
products are in activity equilibrium with parent.

¢ Specific gamma-ray dose constant obtained from Unger and Trubey (1981), gives external photon
dose-equivalent rate in tissue (remv/h) per unit activity (uCi) at distance of 1 m from unshiclded point
source in air; see also Table 6.1.2 of Schleien (1992).

/Mean gamma-ray attenuation coefficient in lead obtained from Unger and Trubey (1981); see also
Table 6.1.2 of Schleien (1992). Reciprocal of value gives thickness of lead (cm) required to reduce
external photon dose-equivalent rate at distance of 1 m from point source in air to 5 % of its unshielded
value.

13
The thermal power listed 1n Table 3.1 1s defined as the energy per unit time emitted by all
1onizing radiations (including the energy of the recoiling decay product nucleus) per unit mass of
the given 1sotope. Thus, shorter-lived radionuclides have relatively high thermal powers, and
longer-lived radionuclides have lower thermal powers. For each shorter-lived decay product, the
thermal power 1s normalized to unit mass of the long-lived parent isotope, taking into account the
branching fraction given in Table 2.1. The thermal power i watts per gram (i.e., joule per second
per gram) can be converted to units of watts per curie by dividing by the specific activity. When
expressed m terms of activity, the thermal power depends only on the energices and intensities of
emitted radiations, but does not depend on the half-life of the radionuclide.

The specific gamma-ray dose constant (I") and mean gamma-ray attenuation coefficient (1) in
lead listed in Table 3.1 are useful indicators of the potential importance of external radiation
exposure. These quantities depend on the energies and intensities of all photons emitted in the
decay of the given isotope. The specific gamma-ray dose constant is defined as the dose-equivalent
rate in tissue per unit activity at a distance of 1 m from an unshielded point source in air. The
conventional unit of dose equivalent is the rem and the SI unit is the sievert, with 1 Sv = 100 rem.
The reciprocal of the mean gamma-ray attenuation coefficient is defined as the thickness of lead
that would be required to reduce the external dose rate at a distance of 1 m from a point source in
air to 5 % of its unshielded value.

The specific gamma-ray dose constants and mean gamma-ray attenuation coefficients in lead
listed in Table 3.1 usually cannot be used to estimate external dose from a finite source containing
the isotopes of concern, because the shielding that would be provided by the source itself is not
taken into account. However, these data are useful indicators of whether external exposure would
be an important concemn for materials containing these isotopes. For example, external exposure is
a much greater concern for “*U and its short-lived decay products than for **U, primarily because
the **Bi and **TI decay products of **U emit high intensities of high-energy photons but “*U
emits only low intensities of lower-energy photons (Kocher 1981). This conclusion is indicated not
only by the much higher specific gamma-ray dose constant for **U, with its short-lived decay
products present in activity equilibrium, compared with the value for Z*U, but also by the much
lower mean gamma-ray attenuation coefficient in lead for the important *'’Bi and **T1 decay
products of ***U compared with the value for “*U. The high attenuation coefficient for **U, and
for several of the other 1sotopes listed in Table 3.1 which emit only low-energy photons, indicates
that self-shielding by a finite source would reduce the external dose by large factors.

3.3 BIOLOGICAL EFFECTS

In addition to potential radiation effects from external exposure, uranium and its compounds
present biological hazards when ingested or inbaled. Chemical toxicity appears as kidney damage
and acute necrotic artenal lesions. Soluble uranium compounds (e.g., uranyl nitrate) are relatively
easily absorbed nto the body, resulting in relatively high organ burdens per unit intake. Inhaled
msoluble compounds have a highly toxic effect to the lungs because of radiation damage (Sax
1968). Some compounds associated with certain forms of uranium can also be toxic (e.g., HF that
often is absorbed on UF, and 1s also a chemical reaction product between UFg and water).

The lack of a “stopper” isotope in the decay chain of 22U results in a dose from ingestion or
PP P Y 2

inhalation about four times greater than the dose from the same activity intake of U (Till 1976).
Table 3.2 lists the concentration limits 1n air and water for releases to the environment which have

14
Table 3.2. Limits on activity concentrations in air and water for releases to the environment for
isotepes of uranium and thorium and for ***Pu and **'Am*

 

Clearance Concentration limit Concentration limit

 

Isotope class ° in air (uCi/mL)° in water (pCi/mL)°

6x 1071 6 x 1078
5x 1071°
1x 107

23 ZU

3 x 10712 3x 107
1 x 1072
5% 107

233U

3 x 10717 3 x 107
1x 1072
5% 107

234U

3 x 10717 3 x 107
1x 1072
6 x 107
236(] 3x 1077 3x 1077
1x 10712
6 % 10—-14

3% 1071? 3% 1077
1x 1072
6 x 107"

238U

2327 4% 107 3x 107

6 x 107

23%py 2% 107 2x 107

2 x 107"

“E <E KU <KEHU <EU0 <K2U <KF2U <=9

=

1 Am 2% 10714 2 x 107¥

 

“Values obtained from Table 2 of 10 CFR Part 20 (U.S. NRC 1991) give limits in air and water for
releases to unrestricted areas which may be accessed by the public. Concentration limits are inversely
proportional to radiation doses per unit activity intake via inhalation (air) or ingestion (water).

? Clearance of inhaled radionuclides from respiratory tract in matter of days (D) for soluble chemical
forms, weeks (W) for chemical forms with intermediate solubility, and vears (Y) for insoluble chemical
forms. Uranium or thorium in insoluble oxide forms should be Class Y. _

“ Corresponding concentration limits in units of pg/mL can be obtained by dividing values in units of
uCi/mL by specific activity of isotope given in Table 3.1.

15
been established by the U.S. Nuclear Regulatory Commission (NRC) (U.S. NRC 1991) n its
radiation protection standards for the public. The limits on activity concentrations in air and water
for releases to unrestricted areas that may be accessed by members of the public as given in

Table 3.2 are based on a limit on an annual commutted effective dose equivalent of 50 mrem

(0.5 mSv) from inhalation and ingestion, respectively. The effective dose equivalent is a weighted
sum of dose equivalents to different organs or tissues defined by the Intemational Commission on
Radiological Protection (ICRP) (1977), and the committed dose 1s the dose received over 50 years
following an acute intake of a radionuclide. For any radionuclide, the committed dose includes the
contributions from any radioactive decay products anising from decay of the radionuclide 1n the
body. For inhaled materials, concentration limits for different lung clearance classes (i.¢.,
solubilitics in the lung) are given. The concentration limits for Class Y materials should be
appropriate for materials contaiming high concentrations of uranium in an nsoluble oxide form.
Any thorium present in the materials also should be Class Y.

The concentration limits in air and water presented in Table 3.2 are inversely proportional to
the internal doses per unit activity intake via inhalation and ingestion, respectively. The dose per
unit activity intake of a radionuclhide is a measure of its radiotoxicity. Thus, the data in this table
indicate that the longer-lived isotopes of uranium (i.c., excluding Z*U) are less radiotoxic than the
shorter-lived ***U, *°Pu or *’ Am, and that all longer-lived isotopes of uranium have essentially the
same radiotoxicity .

The Iimits on activity concentrations in air and water in Table 3.2 may be converted to mass
concentrations in units of pg/mL by dividing by the specific activity of the particular isotope given
in Table 3.1. Thus, for example, the concentration limit for “*U in air for Class Y materials
corresponds to a mass concentration of 5 x 10"? pg/mL., whereas the corresponding mass
concentration for **Uis 3 x 10°* pg/mL. However, for purposes of radiation protection and
radiation dose estimation, the activity of an isotope rather than its mass, usually is the quantity of
mterest.

The quantities of the various radioisotopes i the matenals of concemn to this report are
usually reported in terms of mass. However, as noted previously, the quantity of interest in
addressing radiological concerns, including radiation doses from management and disposal of the
materials and from any accidental releases, usually is the activity of the various isotopes. For any
radioisotope, the mass and corresponding activity are related by the specific activity given in
Table 3.1. In this section, the specific activities of the different isotopes in the matenals given in
Table 3.1 are combined with the radiation doses per unit activity inhaled or ingested, which are
inversely proportional to the concentration limits m air or water in Table 3.2, to estimate the mass
abundance of *U relative to that of the other isotopes above which the **U would be the most
important radiological concern.

The radiological significance of **U relative to ***U is described as follows. The data in
Table 3.2 show that the doses per unit activity inhaled or ingested are essentially the same for **U
and ?°U. Therefore, *U would be radiologically more significant in the materials if the activity of
U exceeds the activity of “’U. Based on the specific activities of the two isotopes given in
Table 3.1, the activity of “*U would be greater than the activity of *’U if its mass abundance
exceeds about 0.02% of the mass abundance of *’U. In other words, 1 g of U has the same
radiological significance as 4400 g *°U.

16
A similar analysis for the other long-lived isotopes of uranium (i.c., excluding **U) gives the
following results: 1 g U has the same radiological significance as 1.5 g Z*U, 150 g #°U, or
28,000 g ®*U. As a further example, consider highly enriched uranium (HEU) with mass
abundances of 93 wt % U, 6 wt % *U, and 1 wt % “**U. This isotopic distribution is typical of
weapons-grade HEU (Uranium Storage Assessment Team 1996). One g of U has the same
radiological equivalence as 150 g of weapons-grade HEU.

All #*U materials contain small amounts of “*U. Because of its high specific activity and the
presence of its shorter-lived, photon-emitting decay products, the activity of *U is often
sufficiently lugh such as to be an important radiclogical concern. However, because of its
relatively short half-life, the “*U is of concem primarily during storage or operations. It should not
be an important concern compared to that of “*U following disposal, especially if the materials
were placed 1n a disposal facility which isolated the waste from the biosphere for several hundred
years or more.

Finally, **U materials also contain some Z*Th. However, based on the very low specific
activity of this isotope compared with the values for U and **U and the other radiological data
given in Tables 3.1 and 3.2, the small amounts of thorium that would be present would not be
radiologically significant.

The analysis previously described shows that for potential radiation effects from ingestion or
inhalation, “?U is the most important isotope radiologically in all of the materials of concern to this
report. Therefore, it 1s reasonable to identify the matenials in terms of this 1sotope.

3.4 NUCLEAR CRITICALITY

Because **U is readily fissile, care must be taken in the design of process equipment and
procedures to avoid criticality. The critical mass of **U varies from a few hundred grams to a few
tens of kilograms depending on density {or concentration), moderation, reflection, geometry {(or
shape), interaction with other fissionable material in an array, and presence of neutron absorbers.
Table 3.3 summarizes the calculated single-parameter limits for metals, oxides, and solutions of
231J and ®°U reflected by an effectively infinite thickness of water (Pruvost and Paxton 1996).

Nuclear criticality of fissile material is controlled through the balance of neutron production
(i.e., through the fission process) with neutron losses (1.e., leakage from the fissile material system
or nonfission neutron capture). Two common approaches to ensuring subcriticality are geometric
spacing of fisstle material, which enhances neutron leakage from the system, and use of neutron
absorbers. Geometrically safe design of equipment 1n a large-capacity processing plant is
expensive. Many different neutron absorbers (boron, gadolinium, cadmium, “*U) are available.
However, nuclear criticality in *U systems can best be avoided by isotopic dilution of the *°U
with the nonfissile neutron absorber “*U. It is important to note that because all uranium isotopes
have the same chemical characteristics, the “>U will not separate from the fissile uranium (which
could be Z*U or “’U) in any normal chemical process, either before or after disposal.

17
Table 3.3. Values of basic nuclear safety parameters’

 

 

 

 

 

 

 

 

Mass of fissile Cylinder Slab
Material and form nuclide, kg diameter, cm | thickness, cm

3 metal 6.0 4.5 0.38
25U metal (5 wt % 2°U) 20.1 7.3 1.30
33U0, with less than 1.5 wt % water 10.1 7.2 0.8
310, with less than 1.5 wt % water 32.3 116 2.9
313,05 with less than 1.5 wt % water 13.4 9.0 1.1
25,05 with less than 1.5 wt % water 44.0 14.6 4.0
2300, with less than 1.5 wt % water 15.2 9.9 1.3
51J0O, with less than 1.5 wt % water 51.2 16.2 46
#3UO,F, solution 0.54 10.5 2.5
Z5U0,F, solution 0.76 13.7 44
“3UO,(NOs), solution 0.55 11.7 3.1
B3j0,(NOs), solution 0.78 14.4 4.9

 

 

 

 

 

“based on Pruvost and Paxton (1996).

General dilution requirements using DU (in this case, 0.2 wt % >°U and 99.8 wt % ~*U) were
developed to ensure the subcriticality of infinite homogeneous mixtures of *U, DU, quartz sand
[silicon dioxide (S10;)], and water (H,0), and of mfinite homogeneous mixtures of uranium
enriched in *U plus DU. Quartz sand and H,O were selected as the most restrictive materials for
subcriticality that are naturally occurning in large process systems and geological environments.
Other neutron-absorbing compounds consisting of 1ron, calcium, and sodium cannot be ensured to
be present in any specific proportion, consequently, they were not considered in this study.

The Standardized Computer Analyses for Licensing Evaluation (SCALE) sofiware and
neutron cross sections (Lockheed Martin Energy Systems 19935) were used to evaluate subcritical
mixtures of these materials. The selected subcritical value for the infinite-medium neutron
multiplication factor (k) for the **U mixtures was &, < 0.95. The limiting subcritical enrichment
(Paxton and Pruvost 1987) for optimally moderated homogeneous aqueous systems 1s well defined
to be 1 wt % U and 99 wt % “*U. The 1 wt % “°U value was used to define the subcritical DU
dilution relationship for uranium enriched in “’U. Using the results of the computational study for
23U dilution and knowledge about the subcriticality of aqueous, homogeneous 1 wt % °U-
enriched uranium, a simple equation was developed to define the necessary DU dilution to ensure
the subcriticality of **U and uranium enriched in **U. The developed relationship for the most
restrictive combinations of *>U, enriched uranium, and DU is based upon the commonly accepted
concept that two or more mixtures of optimally water-moderated, subcritical (1.e., maximum
ke < 1.0), nfinite-medium fissile matenals may be homogeneously combined and remain
subcritical if the composition of the matenals remains homogeneous.

The neutronic computations performed in this study used the SCALE system, AJAX, and
CSAS1X sequence (BONAMI, NITAWL, XSDRN), with the 238-energy group ENDF/B-V
neutron cross-section librarv. The computations were executed on the Oak Ridge National
Laboratory (ORNL) Computational Physics and Enginecring Division Nuclear Engineering
Applications section workstation, CA01.

13
Historic validation studies (Jordan et al. 1986; Primm 1993) using ENDF/B-V neutron cross
sections have demonstrated that water-moderated, homogeneous, single and multiunit ***U critical
systems have calculated &> 0.95 (average k.5~ 0.99). Therefore, the CSAS1X sequence was
executed for various combinations of Si0z, H;0, *U, and DU (i.¢., 0.2 wt % “°U and 99.8 wt %
#5) to calculate a subcritical multiplication factor for an infinite, homogeneous, medium, k..,
approximating 0.95. The use of a k.. acceptance value of 0.95 for this **U scoping study is not
fully justified because integral experimental data for combined Si0; H,0, **U, and **U mixtures
are not available for data testing and validation. Additionally, specific validation and analytical
studies involving the use of configuration-controlled hardware and software relative to thesc
systems and materials are necessary to satisfv criteria for computational safety evaluations.
Obtaining experimental benchmark data is a primary hurdle for researchers before they can
complete such a specific validation.

Because the physical and chemical conditions of U and ’U for some types of process and
disposal options cannot be guaranteed, the results of this isotopic dilution study were reduced to
the most restrictive possible combination of materials (i.e., Si02, H,0, DU, ?°U, *°U, and **0)
that will ensure subcriticality. This approach also ensures criticality control for typical process
systems. As determined from these computational studies, the most restrictive combination of
materials 1s a homogeneous mixture of uranium and water. For this study, the mixture was
assumed to be a mixture of water molecules and uranium atoms.

Under these restrictive conditions, a simple equation was developed to ensure the subcriticality
of **U and uranium enriched in **U by dilution with DU. The equation defines the quantity of DU
that must be blended with **U and various enrichments of *U. The mass of DU is expressed in
terms of *’U and enriched uranium masses as:

E-1
DU=188-g233U+[—6—E—;—J-EU, Eq. (3.1)
where

DU = gof DU (ie, 0.2 wt % ~°U)
E = 100% x g *°Ulg Z*U + g #°1))
EU = goftotal U-g*°U.

It

i

Use of this equation results in a mixture of uranium that contains <1 wt % **’U and
<0.53 wt % 2°U. In Eq. (3.1), ®*U and *U may be considered to be “*U, provided the atom ratio
(U + P°U)*°U does not exceed 1.0. If the calculated quantity of DU [using Eq. (3.1)] is
negative, the uranium material already contains “*U in sufficient quantity to ensure subcriticality,
and no additional DU is needed.

Equation 3.1 is a good first approximation for diluting “*U and enriched uranium, provided
that the mixture is homogeneous and consists of uranium compounds (excluding compounds of
beryllium and deuterium) and water. The presence of other fissionable materials or non-neutron-
absorbing, highly neutron-moderating elements, such as nuclear-grade carbon, beryllium, or
deuterium, has not been considered in this work. Though other scattering or absorbing nuclides
may be present in a mixture, their effects have not been accounted for in estimating the required
DU mass for dilution of **U and enriched uranium.

19
Because the dilution equation uses DU as the diluent to approximate an equivalent 1 wt %
U enriched uranium and water-moderated system, the potential for an autocatalytic criticality
accident (Kastenberg et al. 1996) is rendered impossible, because homogencous systems of 1 wt %
23U or ~0.53 wt % U cannot be made critical as a mixture of U and H;0.

20
4. STORAGE REQUIREMENTS

The characteristics described in the previous two sections indicate that “*U requires special
considerations when it is being stored or handled. The requirements for *’U-bearing material are
generally much stricter than those for HEU or Pu. This section will describe six basic
requirements: material form, packaging, confinement, criticality control, shiclding, and safeguards.
Each of these is based on a characteristic of “°U: chemical form for long-term material stability,
packaging for reliable containment, safeguards because of its potential use in weapons, criticality
control because of its fissionability, ventilation because of the formation of radiologically or
chemically toxic volatiles in the decay chain or by radiolysis, and shielding because of its
radiological properties. The objective will be to state the requirements, to compare and contrast
them to HEU and Pu, and to show the effect of 1sotopic dilution on the requirements. (Because of
the chemically identical nature of all uranium isotopes, the elemental term, uranium, will be used
where chemical charactenstics prevail).

4.1 MATERIAL FORM

For long-term storage, U must be in a stable form that poses minimal impact on
containment and criticabty control. The overwhelmingly preferred form for long-term storage is
Us0; (Cox 1995). It is the most chemically stable form under normal storage conditions, and it
acquires the fewest waters of hydration (1.e., moderators) per uranium mass of any uranium
compound.

Other forms of uranium may be acceptable for certain storage conditions (e.g., shorter storage
periods, inert atmospheres, and special packaging forms such as clad fuels). Metal has the
advantage of being very compact (with extremely high density) when stored as large billets and
having no waters of hydration. However, its metallic nature can only be relied upon if its surfaces
are protected from oxidizing atmospheres to prevent any conversion to oxides (with a resultant
volume increase and powdery surface) and a tendency to acquire commensurate waters of
hydration.

Additionally, because of the high specific activity of **U, (which for pure ***U is about one
order of magnitude lower than **Pu, but can exceed that activity with higher levels of U
contaminant) contaminants must be kept to a mmimum. Contamimants also must be factored into
storage considerations because they tend to evolve to gases as they absorb the energies of
radioactive decay emissions. For example, **U fluorides tend to evolve fluorine gas that can
pressurize the storage container or contaminate the storage arca atmosphere with its toxic and
corrosive by-product, hydrofluoric acid. A greater problem is water contamination (hydration and
adsorbed waters) which decomposes radiolvtically to form hvdrogen and oxygen gas. Hydrogen
gas is not only a container pressurization problem, but also a potentially explosive-mixture
problem. In a similar manner, plastics used in packaging may also decompose radiolytically to
generate hydrogen (and carbon mono- and dioxides) that, when combined with air, also can create
a potentially explosive atmosphere. Care must be exercised 1n treating these radiolytic gases to
prevent any toxic, corrosive, and explosive consequences.

21
4.2 PACKAGING

The packaging material for “°U must reflect compatibility with (a) the chemical nature of the
contained material and the storage atmosphere and (b) the high specific activity of the stored
radionuclides. Similar to Pu, the high specific activity of *’U and its associated isotope, ~*U,
essentially eliminates using organic materials for primary container construction, except during
brief periods of storage. Because of radiolytic degradation of hydrocarbons, even organic
materials in gasketing and bag-out layers (i.¢., outside the primary can) cannot be relied upon for
maintaiming their integrity and therefore also must be considered a source of radiolytic gas
generation. ‘

The contaner’s materal of construction must also be compatible (1.¢., resist corrosion upon
contact) with the contained material and with the storage environment. For example, type 304L
stainless steel provides a robust barrier 1n the absence of chlondes (except as trace impurities) both
in the contained material and in the storage atmosphere. Similarly, aluminum provides a reliable
barrier in the absence of nitrates. Multiple lavers of packaging can be made of different materials
of construction to address complex compatibility issues.

Containers of “*U may be closed such that they provide a gas-tight seal for, or allow venting
of, evolved gases. A seal is usually achieved by welding, although other closures are possible.
Sealed containers trap evolved gases up to a pressure at which the container fails and vents to its
surroundings. Sealed containers containing *U with water or plastic trap evolved hydrogen.
Therefore, to avoid a hydrogen explosion or deflagration, particular care must be exercised when
handling and opening containers that have beeri sealed for an extended period of time.
Consequently, it is highly desirable that materials to be stored are dnied before the containers are
scaled. Containers that allow venting of evolved gases either have designed, filtered vents, or they
allow leakage of such gases through imperfect closures (e.g., screw-top lids). If gas leakage1s a
possibility (either via imperfect closures or as a result of a sealed contamner failure), then
ventilation/confinement systems (discussed later) are warranted.

4.3 CONFINEMENT

On a unit mass basis, the alpha activity of °U is one order of magnitude less than that of
#°Pu, but is three orders of magnitude greater than that of “’U. Thus, the alpha decay
characteristics (and, thercfore, the confinement requirements) for “>U are between the other two
strategic fissile materials. The presence of “*U contributes additional alpha activity which can
cause the activity of **U-bearing material to exceed that of “’Pu. Containment and ventilation are
nsed to protect workers and the public from inhalation exposure to “**U by maintaining radiological
confinement of designated areas. Confinement provides a physical barrier and ventilation draws
air from areas of lower radiological contamination to arcas of higher contamination. Additionally,
ventilation provides a means of filtration before atmosphenic discharge.

The high specific activity of “°U (and *U) promotes evolution of gaseous decomposition by-
products from contaminants (such as water and plastics). These by-products include potentially
explosive components (e.g., hydrogen) which must be 1solated or carned away by a rehiable
ventilation system.

22
Material confinement is also challenged as alpha particles (with a2 mass of 4 atomic units) are
gjected at high energies from their immediate parent nuclides (with masses ranging from 212 to 232
atomic units). The parent nuclides react to the alpha particles’ “pushing off” by recoiling to this
force. This phenomenon is known as “alpha recoil.” In matenals with significant concentrations
of **U (tens of ppm or higher), the net effect is a slight migration of radioactive particles. A
properly designed confinement and ventilation system prohibits migration particles to areas of
Iower contamination and sweeps them into areas of higher contamination or to filters. The
ventilation system also can become important for dissipating the heat generated by these highly
energetic alpha emissions.

Unlike the decay chains for “°U or Pu, one of the products in the **U decay chain includes a
radioactive inert gas, “°Rn, with a 55.6 s half-life. Any decay product immediately dissociates
from any compound to which its parent was bound. However, since Rn is inert, it will not form a
new bond after dissociation. Furthermore, because 1t 15 a gas, it will join the other gases in its
environment and can pass unhindered through particulate filter media. Therefore, the presence of
significant amounts of “**Rn requires retention of the off-gas to allow this isotope to decay through
several half-lives to ensure that it has transmuted to the filterable decay product, *°Po. Itis
important to note that “°Rn retention is more crucial at higher “*U concentrations (10 ppm or
greater) and for conditions during which purging of evolved radon is facilitated by gas flow
through the bulk material (which is more typical during processing such as air-sparging of liquids).
However, even for dormant storage, “’Rn evolution must be considered in ventilation system
design.

Similar arguments may be made for ventilation in the example of “°U stored as UF,, which
has a significant vapor pressure. Allowing this material to escape containment would be hazardous
not only radiologically but also chemcally because of the release of HF and F,. Both of these
latter matenals are toxic and corrosive to ventilation and filtration equipment that is not
particularly suited for this service.

The requirement for ventilation becomes less crucial with high-integnty packages such as
those considered to be “special form™ (49 CFR Part 173) with two corrosion-resistant, certified-
welded layers of metal containers. As long as their integrity is intact, such special form canisters
can be considered reliable for containing alpha particles, recoiling parent particles, radon, and even
radiolytically generated gases.

4.4 CRITICALITY CONTROL

Because “U is readily fissile, care must be taken in the design of storage facilities to avoid
criticality. DOE requires adherence to the American National Standards Institute, Inc./American
Nuclear Society (ANSI/ANS) nuclear criticality safety standards (U.S. DOE 1992a;
ANSI/ANS-8.1 1983).

Nuclear criticality of all fissile material is controlled through the balance of neutron
production (i.e., through the fission process) with neutron losses (1.e., leakage from the fissile
material system or nonfission neutron capture). This ratio of neutron production to neutron loss
(k. must be kept less than 1 under all circumstances. Typically, k.45 1s kept less than 0.95.

23
Two common approaches to ensuring subcriticality are geometric spacing of fissile material
and using neutron absorbers. Geometrically safe configurations enhance neutron leakage from the
system, although such designs may be expensive. Many different neutron absorbers (boron,
gadolinium, cadmium, 2*U) are available. It is important to note that because all uranium isotopes
have the same chemical characteristics, the ***U will not separate from the fissile uranium (which
could be *U or ***U) in any normal chemical process.

Ideally, the k. should be established based on experiments with an allowance for uncertainty.
In practice, especially for *U, there is a lack of experimental data. In such cases, calculational
methods, as described in Section 3.4, may be used. Bias is then established by correlating the
calculations to known experimental results. Trends in the bias are used to extend beyond the range
of experniments.

A key concept in criticality control is the double contingency principle. This principle, which
1s a safety factor that 1s built mnto storage design, requires at least two independent changes in
system conditions before a cniticality accident can occur.

In addition to the technical practices previously descnibed, administrative controls should also
be established to prevent accidental criticality. These may include process analyses, matenal
control, emergency procedures, and operational control and review. Most of these factors take
added importance in processing facilities at which factors such as their gcometries mayv be subject
to change.

4.5 SHIELDING

Uranium-233 compounds often must be handled in shielded enclosures because of the high
external radiation hazard (Horton 1972). The external radiation field for **U depends on many
factors. Among these are the surface area of the source and the distance from the source. Self-
shielding because of the density and geometry of the material is another factor. Finally, external
shielding can be used to reduce the field.

While self-shielding and stainless steel containers provide a small reduction in the external
radiation field, the primary shielding materials used to protect workers and the public are lead and
concrete. Again, the amount of material needed depends primarily on the amount of Z*U present.
Figure 4.1 shows the necessary lead shielding to reduce the dose from 1 kg of 35-d aged *U to
2 mR/h at a distance of 1 m (Amold 1962).

Diluting the U with DU may help reduce the shield thickness requirements. The added

uranium mass will provide an additional degree of self-shielding, but will pose a substantially
larger mass (and volume) to be shielded, which will increase the overall shielding mass.

24
4

Shielding (cm of Pb)

ORNL DWG 98-3898

 

18

16 4

14 -

12 -

10 -

 

14 §

 

 

 

50

;1 1 2
. v ¥ . s %

160 150 200 250 300 350
B2 concentration (ppm)

Fig. 4.1. Lead shielding for *** w2

U with various concentrations of = U,

400

450

500
4.6 SAFEGUARDS

Because of its fissile nature, “*U may be used to produce nuclear weapons. Therefore,
safeguards to prevent theft must be applied. Currently, DOE requirements are used exclusively for
the nation’s **U inventory (U.S. DOE 1994). However, the United States is under international
treaty obligations which could place the **U under International Atomic Energy Agency (IAEA)
safeguards as well (U.S. DOE 1992b). The requirements of these two organizations are similar.
Both organizations use a graded approach to safeguards in which material that 1s most effective in
making nuclear weapons is placed under the greatest control. Table 4.1 summarizes the different
levels of the DOE requirements. IAEA requirements, which correspond roughly with DOE
Attractiveness Level B, are also given for companson. The specifics will be discussed in further
detail in the following sections.

Table 4.1. DOE nuclear material safeguards categories

 

 

 

 

 

 

 

 

 

Categorv (quantities in kg of “*U)
Attractiveness level I I I v
(Highest safeguards) (Lowest safeguards)
A (most attractive) All quantities NA® NA NA
B =2 204 10<2 202t0<04 <0.2
(IAEA)® (>2) (> 0.5 to <2) (£0.5) (NA)
C >6 >22t0<6 204 10<2 <0.4
D NA >16 >31t0<16 <3
E (least attractive) NA NA NA Reportable
quantities

 

 

 

 

 

 

"NA-—-Not applicable
*IAEA values are included for comparison

4.6.1 DOE Requirements

Under DOE Orders, 2°U is separated into four categories according to the amount of material
present and its attractiveness level. The attractiveness levels correspond to the case in which the
matenal can be used i creating nuclear weapons.  The most attractive matenals (Level A) are
assembled weapons and test devices. All quantities of Level A fall into Category 1. Pure products
{c.g., pits, major components, buttons, ingots, recastable metal, and directly convertible materials)
form Level B. These fall into categories I-IV according to the amount of material. High-grade
material [e.g., carbides, oxides, solutions (= 25 g/L), nitrates, fuel elements and assemblies, alloys,
and mixtures] fall into Level C, which is also further separated into four categories according to
the amount of matenial. Level D consists of low-grade matenals (e.g., solutions with 1-25 g/L or
process residues that require extensive reprocessing). These materials are classified as only
Category I, III, or IV. The lowest level of attractiveness (Level E) consists of gram quantities or
greater of uranium existing as highly irradiated forms and solutions (U.S. DOE 1994).

It should be noted that these categories make no distinction as to the isotopic concentration of

**U. This is in sharp contrast to **U. Material enriched to 50 wt % “>°U or greater falls into
attractiveness Level C (or greater). Material enriched to greater than or equal to 20 wt % and less

26
than 50 wt % in *°U falls into Level D. Less than 20 wt % enrichment in *°U belongs to Level E.
Because *U was not originally deployed in nuclear weapons or commercial nuclear power plants,
safeguards requirements as a function of 1sotopic levels have not been developed.

There are three functions of material control: access controls, material surveillance, and
material containment. Each of these functions also takes a graded approach based on the category.
Access controls are concerned with preventing unauthorized personnel access to materials, data,
and equipment. The graded approach ranges from simple admimstrative controls for Category IV
matenal to extensive, complex procedures for Category I material. Access controls are also
designed to prevent Category 1l and IV materials of Levels B or C from accumulating into
Category I or Il amounts. Finally, there is a performance requirement that tests to detect
unauthorized access to Category I or II matenal be at least 95% effective.

Material surveillance has as its goal the detection of unauthorized flows of materials out of the
material containment areas. This goal 1s accomplished using sensors, patrols, logs, tamper
indication devices (TIDs), portal monitoring, waste monitoring, and other administrative checks.
As with matenial access, the performance requirement for Category I and I material is that
unauthorized actions must be detected in at least 95 % of tests. Surveillance ensures that
Categories | and I maternals are only used in the authorized locations described below. Category
I materials that are outside of locked storage areas are also required to be kept under surveillance
within authorized areas. The requirements for Category IV material are site-specific.

Material containment 1s broken into three areas subject to access controls and surveillance.
The Materials Balance Area is the general term for any area i which nuclear materal 1s used,
processed, or stored. In accordance with the graded approach, the Protected Area, which is used
for Category Il matenials, has stricter access controls and increased surveillance. Finally, within
the Prorected Area is the Materials Access Area, in which Category I matenal is used, processed,
or stored.

4.6.2 JAEA Requirements

IAEA 1s the branch of the United Nations concerned with controlling the spread of nuclear
weapons. While the responsibility of protecting fissionable material lies with individual nations,
IAEA 1s authorized under 1its statute to establish and administer nuclear matenials safeguards.
These safeguards are designed to verify that nuclear materials and other nuclear-related items are
not used to further any military purpose. IAEA also applies safeguards (at the request of a nation)
to any of that nation's atomic energy activities (IAEA 1956).

Currently, about 1000 U.S. sites and facilities are eligible for JAEA safeguards. This means
that the U.S. government has declared that the site is not critical to national security and that JAEA
has the right to pick the site for implementation of safeguards. However, because of its limited
budget, it has generally been IAEA’s policy to limut its selection of sites in declared nuclear
weapons states. In this way, it can concentrate on preventing the further spread of nuclear
weapons to nonnuclear states. Only three sites (vaults at Y-12, Hanford, and Rocky Flats) in the
United States have some of their inventories under JAEA safeguards. All three were selected by
voluntary offer agreements for which, for political reasons, the U.S. government paid IAEA to
implement safeguards. IAEA is also overseeing the downblending of Russian matenal at the
Savannah River Site (SRS) and Kazak material at Babcock and Wilcox.

27
The first IAEA activity 1s verification of storage design. For this phase, a storage facility
would complete IAEA’s Design Information Questionnaire (DIQ). Typically, there is a two-month
period after selection before the DIQ is due.

The second stage of the process 1s verification of the stored quantity of nuclear material.
Verification is accomplished by measuring items via sampling by destructive assay on a small
selection of random 1tems and nondestructive assay of a larger fraction of the items. IAEA then
places the items under containment/surveillance (C/S) using cameras and TIDs.

Future inspection and mventory activities depend on the designation of the storage area and
the safeguards approach applied by IAEA. At the worst extreme, future inspections (twice a
month) would verify a random sampling of TIDs and perform gamma spectrometry verification of
a random sampling of items. During an annual physical inventory, a random sampling of items
would be removed for nondestructive measurements. The option of opening containers and
removing samples for destructive measurements is reserved by IAEA. Because of the extreme
gamma radiation hazard of “*U (due to the presence of **U), significant handling precautions and
expenses would be incurred.

Currently no *U is under IAEA safeguards. However, JAEA does make recommendations
on the physical protection of “’U (IAEA 1980; IAEA 1989). These recommendations depend on
the following categorizations according to mass: 2 kg or more of unirradiated ***U is Category I,
between 500 g and 2 kg 1s Category I, and 500 g or less 1s Category III. Radiologically
insignificant quantities and irradiated ’U are exempted from these classifications. In contrast, as
with DOE regulations, **U is categorized not only by mass but also according to three levels of
enrichment: greater than 20%, 10-20%, and up to 10%. Similar to DOE regulations, JAEA
safeguards do not account for different isotopic levels of *U. The limited use of **U to date has
not warranted development of such safeguards regulations.

The recommendations of IAEA for protecting materials have certain concepts which are
generic to all three categories. Matenals need to be stored in arcas to which access 1s controlled.
All personnel working in the facility need to be trained about the importance of physical protection
and the appropnate responses in cases of emergency. Alarms and guards should be used to detect
and respond 1o sabotage or unauthorized removals of matenal. Finally, a security survey should be
made whenever a significant change mn a facility or its function takes place. This survey is a
critical examination to evaluate, approve, and specify physical protection measures.

As in DOE requirements, storage of Categories 1 and I nuclear matenals requires a protected
arca. This arca is under constant surveillance, either by guards or electronically, and is surrounded
by a physical barrier. Access to this area 1s kept to the minimum necessary and controlled through
a Iimited number of entry points.

Category I maternials are isolated further in an mmner area within the protected arca. (This
corresponds to the Marerials Access Area of DOE requirements). This inner area 1s arranged with
a minimum number of alarmed entrances and exits (ideally, only one). The storage area itself
should be alarmed and locked. Authority as to who has these kevs should be tightly controlled.
Electronic surveillance should be effected using at least two independent transmissions.

28
4.6.3 Elimination of Weapons Potential

The surest way to safeguard **U from theft or misuse is to eliminate its ability to be used for
weapons. Studies indicate that U can be made unsuitable for military use by diluting it with DU
to a fissile concentration of 12 wt % (Dubrin 1995; Benedict et al. 1981). This level of dilution is
equivalent to diluting weapons-grade HEU with **U to 20 wt % “°U.

Earlier studies on demilitanization of the large quantities of weapons-grade HEU indicate that
isotopic dilution of **U with Z*U is the preferred demilitarization option (U.S. DOE 1996a). The
U.S. government has issued a Record of Decision (U.S. DOE 1996b) making isotopic dilution the
official policy for demilitarization of HEU. Given the relatively low cost, assured technical
feasibility, and acceptance for demilitarization of HEU, the same approach may be used for
demilitarization of *’U. It is noted that any “°U inventory that contains sufficient *°U (>20 wt %
) is classified as HEU; therefore, isotopic dilution is the stated policy for demilitarization of
this matenial. As with safeguards categories, the required levels of isotopic dilution to eliminate
weapons potential for *U have not been implemented in DOE orders or IAEA regulations.

29
5. DISPOSAL REQUIREMENTS FOR #°U DECLARED AS WASTE

Current legal and regulatory requirements for disposal of waste materials containing 2*U
are discussed in this section. Most requirements for disposal of any waste containing “>U depend
on the activity concentrations of that isotope and the activity concentrations of other radionuclides
that are present. In addition, if waste materials exhibit certain chemical characteristics or contain
particular hazardous chemicals, then disposal must comply with requirements for solid hazardous
waste developed under authority of the Resource Conservation and Recovery Act (RCRA).

5.1 RADIOACTIVE WASTE CLASSIFICATION

All waste materials containing “*U of concem to this report were produced as a result of
operations of nuclear reactors. Therefore, because the use of nuclear reactors is authorized only
under the Atomic Energy Act (AEA), these radioactive materials would be managed under
authority of the Act. Furthermore, DOE currently is responsible for management of these
materials because they were generated at DOE sites or under DOE agreements. The different
classes of radioactive waste anising from operations of the nuclear fuel cycle that have been defined
in law include SNF, HLW, TRU waste, LLW, and uranium or thorium mill taithngs.

Any waste materials containing “*U and any wastes arising from further processing of these
materials would be classified as LLW, regardless of the concentrations of “°U or any other
radionuchides. This classification 1s based on the current legal definitions of LLW and the fact that
the materials containing mainly **U and any wastes arising from further processing of these
materials do not meet the classification criteria for HLW or TRU waste (as well as SNF or mill
tailings).

LLW is currently defined only by exclusion as waste that 1s not SNF, HLW, TRU waste, or
uranium or thorium mill tailings. Furthermore, the current legal defimitions do not describe the
constituents or properties of LLW. Also, in contrast to the definitions of HLW and TRU waste,
the definition is not related in any way to requirements for safe handling, storage, or disposal.
While most LLW contains relatively low concentrations of any radionuclides and is expected to be
acceptable for near-surface disposal, LLW also can contain high concentrations of longer- and
shorter-lived radionuchides and require a disposal system similar to that for HLW or TRU waste.

It also should be noted that LLW does not include naturally occurring radioactive matenial other
than source material or radioactive material produced in an accelerator, because these types of non-
fuel-cycle wastes are not subject to federal regulation under authority of the AEA.

HLW is defined as the waste stream that directly results from the chemical reprocessing of
SNF. Therefore, waste materials containing mainly **U would not be classified as HLW even if
they resembled HLW (a) in having high levels of decay heat and external radiation, due primarily
to the concentrations of shorter-lived radionuclides (i.e.,”*U), and (b) in having high concentrations
of long-lived alpha-emitting radionuclides, because they are not wastes which arise directly from
chemical reprocessing of SNF.

By definition, TRU waste contains > 100 nCv/g of alpha-emitting TRU radionuclides with

half-lives > 20 years. Therefore, waste materials containing mainly 2*U would not be classified as
TRU waste, ¢ven though they resembled TRU waste in having high concentrations of long-lived,

31
alpha-emutting radionuclides, because the concentrations of alpha-emutting TRU radionuclides with
half-lives greater than 20 years would be less than 100 nCv/g.

Potentially, waste materials containing mainly *°U could be classifiable as HLW if the
definition in Clause (B) of the Nuclear Waste Policy Act NWPA) were implemented by the NRC
to include waste other than the primary wastes from chemical reprocessing of SNF. Such a
reclassification would require a petition to the NRC. Alternatively, the maternals and wastes
arising from further processing of these matenals possibly could be classifiable as TRU waste if
the definition of TRU waste 1in the Waste Isolation Pilot Plant (WIPP) Land Withdrawal Act were
modified to include waste with high concentrations of alpha-emtting, non-TRU radionuclides.
Although either of these classifications of wastes containing >°U could be reasonable, based on the
radiological properties of the waste, they require changes in regulations (to classify as HLW) or
changes in law (to define as TRU waste).

5.2 CLASSIFICATION AND DISPOSAL AS HAZARDOUS WASTE

Radioactive wastes regulated under authority of the AEA may also be regulated as sohd
hazardous waste under authority of RCRA. Indeed, current DOE policy specifies that all of its
radioactive wastes shall be managed in accordance with requirements of the AEA and RCRA,
unless the waste is shown not to be hazardous as defined under RCRA (U.S. DOE 1988).
However, “*U waste should not be classified as hazardous under RCRA because (a) the materials
are not ignitable, corrosive, or reactive as defined under RCRA, or these characteristics result from
the chemical form of the uranium, which is exempted from regulation under RCRA; (b) any
cadmium in the materials is serving its intended purpose as a neutron absorber to prevent nuclear
criticality and, thus, is not classified as a waste under RCRA, and the materials should not contain
significant quantities of any other hazardous substances of concern to the toxicity characteristic;
and (c) the materials should not contain any listed hazardous wastes.

This conclusion also should apply to any wastes arising from further processing of
Z3U-bearing waste materials. However, it may be necessary to reconsider, on a case-by-case basis,
the question of whether any such wastes contamning cadmium would be considered hazardous under
RCRA, especially if the cadmium were no longer serving its intended purpose as a neutron
absorber to prevent nuclear criticality.

Should wastes containing ~°U be classified as hazardous under RCRA, the conclusions
regarding treatment and disposal requirements may be summarized as follows:

e  The type of disposal system that would be required (e.g., a near-surface facility or
geologic reposttory), based on radiological properties of the waste, would not be affected.

 

" The current legal definition of high-level radioactive waste is contained in the NWPA, and the definition
has two paris as follows:

(a) the highly radioactive material resulting from the reprocessing of SNF, including liquid waste
produced directly in reprocessing and any solid material derived from such liquid waste that contains fission
products in sufficient concentrations, and

(b) other highly radioactive material that the NRC, consistent with existing law, determines by rule
requires permanent isolation.
e If wastes containing “°U were classified as hazardous because they are ignitable,
corrosive, or reactive, as defined under RCRA, due to the presence of substances other
than radionuclides (i.e., uranium), the wastes would need to be treated to remove the
hazardous charactenstics prior to disposal. Afier treatment, the wastes would no longer
be classified as hazardous under RCRA.

e  If wastes containing *U were classified as hazardous because of the presence of
cadmium, the wastes would need to be treated to remove the toxicity characteristic before
disposal of the waste, unless it can reasonably be demonstrated that there will be no
migration of hazardous constituents from disposal units for as long as the waste remains
hazardous. Afier treatment, the wastes would no longer be classified as hazardous under
RCRA.

e If any listed hazardous wastes were introduced into wastes containing 2°U (e.g.,
during processing), the wastes also would need to be treated prior to disposal (unless
the condition of no migration from disposal units can reasonably be demonstrated), but
the treated wastes still would be classified as hazardous under RCRA.

As noted previously, waste which 1s ignitable, corrosive, or reactive due only to the presence
of uranium would not be classified as a hazardous waste under RCRA and, thus, would not require
treatment under RCRA to remove any such hazardous characteristics. However, the waste
acceptance cniteria (WAC) for any potential disposal facility presumably would require that the
uranium not be in a chemical form which is ignitable, corrosive, or reactive, in which case the
uranium would need to be converted to a suitable chemical form {e.g., oxide) before disposal.

5.3 RADIOACTIVE WASTE DISPOSAL OPTIONS

Based on the conclusion that waste materials containing “*U and any wastes arising from
further processing of these matenals would be classified as LLW, a number of options for disposal
are available which depend essentially on the concentrations of 2’U. LLW containing mainly “*U
generally would be intended for disposal 1n a near-surface facility at the generating site. If the
waste would not meet the acceptance criteria for such a facility or if the generating site does not
have such a facility, it would be disposed of at another DOE site with a near-surface facility that
could accept the waste. However, wastes containing high concentrations of *’U probably would
not be acceptable for near-surface disposal at any DOE site. The disposal criteria and the attendant
options are discussed below.

5.3.1 Near-Surface Disposal of LLW

DOE is responsible for disposal of its own LLW under authornity of the AEA. The Low-Level
Radioactive Waste Policy Amendments Act is concerned only with commercial LLW and, thus,
does not apply to DOE's LLW unless it 1s sent to a commercial disposal facility, an alternative
which is not precluded by existing law or regulations.

Most LLW is intended for disposal in near-surface facilities. Current DOE policy encourages

disposal of 1its LLW at the major waste-generating sites (U.S. DOE 1988): Oak Ridge, SRS, Idaho,

33
Hanford, Nevada, and Los Alamos. These sites also may receive LLW from other DOE sites
without disposal facilities or waste that is not acceptable for disposal in a facility at the generating
site. For example, below-grade disposal of any waste was phased out at ORNL in the early 1990s
(Lockheed Martin Energy Research 1997). However, as described below, LLW containing high
concentrations of long-lived radionuclides may not be acceptable for near-surface disposal at any
DOE site.

Currently, near-surface disposal of DOE's LLW is regulated only according to requirements
in Chapter III of DOE Order 5820.2A (U.S. DOE 1988). The NRC's licensing criteria for near-
surface disposal of radioactive waste, as stated in 10 CFR Part 61, which are intended to apply
primarily to commercial LLW, do not apply to DOE's LLW unless it is sent to a commercial
disposal facility. In addition to performance objectives for limiting radiation dose o individual
members of the public, current DOE requirements include a performance objective for limiting
dose to inadvertent intruders onto disposal sites after loss of active institutional control at
100 years after disposal (U.S. DOE 1988). The performance objectives in the DOE Order are
used to establish WACs for near-surface disposal facilities in the form of concentration limits of
radionuchides in waste.

For example, for *’U, *’U, and ***U, an analysis for a near-surface disposal facility at SRS
showed that the limits on average uranium concentrations over the disposal site are on the order of
1-5 nCr/g (Lockheed Martin Energy Systems 1994), and the concentration limits for individual
waste packages might be ten times higher if the limits on average concentrations over the site are to
be met (Wood et al. 1994). The concentration limits for U would be about an order of
magnitude higher. Although such concentration limits are established on a site-specific basis and,
thus, would not be expected to be the same at all sites, similar results should be obtained for near-
surface disposal facilities at most DOE sites other than SRS.

The analysis described above suggests that LLW containing concentrations of > U, *°U, and
%V substantially less than 100 nCi/g could be unacceptable for near-surface disposal at most
DOE sites. This conclusion is consistent with a statement by the NRC that near-surface disposal
of large quantities of DU that would be generated at a commercial uranium enrichment facility is
not appropriate. The expected concentrations of **U in these wastes are as high as 60 nCi/g (U.S.
NRC 1994). Even though the DU would be Class-A waste, as defined by the NRC n
10 CFR Part 61, and Class-A waste normally 1s acceptable for near-surface disposal with minimal
requirements, disposal of large quantities of DU in any concentrations was not contemplated by the
NRC in the development of 10 CFR Part 61. In particular, near-surface disposal of these wastes
may not ensure protection of future inadvertent intruders onto disposal sites, as required by
10 CFR Part 61.

5.3.2 Alternatives to Near-Surface Disposal

For DOE LLW with concentrations of radionuclides sufficiently high that it would not be
acceptable for near-surface disposal at the generating site or at any other DOE site, an alternative
disposal facility would be required. The following two alternatives could be considered.

The first alternative would be to consider use of the Greater Confinement Disposal (GCD)
facility at the Nevada Test Site (NTS). This facility uses deep, lined boreholes that emplace waste
well below the ground surface but far above the water table. In the past, the GCD facility has
disposed of igh-activity LLW and some TRU waste. Therefore, disposal of waste with high

34
concentrations of U presumably would be acceptable, although a WAC for *U wastes has not
been developed for this facility.

However, a disadvantage of this alternative 1s that the GCD facility is not currently disposing
of waste. Disposal activities were terminated following a finding by the State of Nevada that
disposal of radicactive waste in the GCD boreholes constitutes a form of underground injection in
wells that is prohibited by regulations developed by the U.S. Environmental Protection Agency
(EPA) under authority of the Safe Drinking Water Act. Therefore, disposal of additional waste in
the GCD facility could occur only if the finding by the State of Nevada were overturned. |

Other facilities located well below the ground surface are being considered at the NTS for
disposal of high-activity LLW that is not acceptable for near-surface disposal, including disposal
in surface craters produced by underground testing of nuclear weapons. However, no such
facilities have vet been developed.

If it would not be acceptable to dispose of high-activity LLW in any of the facilities at the
NTS described above, the only remaining current altemative is disposal with SNF and HLW ina
geologic repository. This alternative would be consistent with the NRC's current requirements for
disposal of commercial LLW 1n 10 CFR Part 61. These requirements specify that waste that is not
generally acceptable for near-surface disposal-the so-called greater-than-Class-C waste—
requires disposal in a geologic repository, unless disposal elsewhere is approved by the NRC on a
case-by-case basis (U.S. NRC 1989). However, the particular environmental standards, licensing
criteria, and WACs that would apply to disposal of high-activity LLW in a geologic repository are
currently unknown.

5.4 MIXING OF U WITH OTHER WASTES

The disposal options discussed above assume that the materials containing mainly “*U and
any wastes arising from further processing of these materials would be managed and disposed of
separately from other wastes. However, alternatives involving mixing of these wastes with HLW
or TRU wastes before disposal could be considered, particularly if the materals would not be
acceptable for disposal in any existing facilities for LLW at DOE sites.

There are no legal or regulatory prohibitions to converting solid LLW contaimng high
concentrations of *°U to liquid form and mixing the liquid wastes with the much larger volumes of
HLW at the SRS or Hanford sttes. In addition, the resulting wastes should not present any
technical challenges for safe disposal that would not be considered for HLW 1n its present form.
The resulting wastes would be managed as HLW, based on the classification of its most important
component, and would be intended for disposal in a geologic repository.

There also are no legal or regulatory prohibitions to mixing LLW containing high
concentrations of *’U with TRU wastes. If the resulting wastes would be classified as TRU waste,
the definition of which does not depend on the concentrations of any non-TRU radionuclides, the
wastes would be intended for disposal in the WIPP facility, provided the WACs for that facility are
met. As in the altemative of mixing with HLW described previously, the resulting wastes should
not present any technical challenges for safe disposal that would not be considered for TRU waste
in its current form. However, there may be other issues that would need to be addressed.

35
Alternatives for disposal of materials containing mainly U and other wastes containing high
concentrations of > U that would involve mixing with HLW or TRU wastes would need to be
justified compared with managing and disposing of these matenals separately from other wastes.
Mixing with other wastes could be justified, for example, if it would allow more timely disposal in
existing or planned facilities, or it would result in substantial cost savings for waste management
and disposal.

5.5 NUCLEAR CRITICALITY
5.5.1 Need to Avoid Criticality

Nuclear criticality must be avoided in any disposal facility because of the increased potential
for releases of radionuclides to the environment. Evidence from nuclear reactors occurring
naturally in the geological past (Cowen 1976) indicates that such events have generated both added
radroactivity and heat over time periods of hundreds of thousands of years. The heat generated
creates higher disposal site temperatures that (a) drive chemical reactions which, in turn, degrade
waste packages and waste forms; (b) cause water movement within a disposal site that may
transport radioactivity to the environment (Buscheck and Nitao 1993) and (c) contribute to major
uncertanties in site performance.

Nuclear criticality control must also be mamtained in the facilities used to store and process
3 material into an acceptable waste form. The potential quantities of ***U requiring disposition
are small, and some of the *’U contains “*U and its highly radioactive decay products sufficient to
require hot-cell processing.

Existing waste processing facilities may be able to process °U. These include existing DOE
HLW vitrification facilities and proposed TRU waste processing facilities. However, these
facilities are not designed for significant concentrations of fissile materials. If such facilities are to
be used, nuclear criticality safety controls for “*U cannot rely on geometry or chemical control
without substantially changing plant equipment and operations.

NRC crticality control requirements for HLW disposal are specified in 10 CFR Part 60.
These requirements are similar to those for storage. The double-contingency standard is required.
Additionally, the calculated effective multiplication factor, ks, now must be less than 0.95 instead
of less than 1 for storage.

Although elements found in geological deposits may be effective neutron scatterers (e.g.,
silicon, aluminum, or oxygen) or somewhat effective neutron absorbers (e.g., ron, sodium, or
calcium), no assurance can be provided that such elements will remain with the uranium over
geological time spans. Several mechanisms cause changes in waste geometry and chemustry over
geological time frames. These include groundwater transport of uranium and mechanical
disturbances of the waste. If criticality control is to be ensured for thousands of years by cither
geometric control or chemical control (including neutron absorbers), system performance must be
predicted for these lengths of time. Such prediction is difficult.

36
5.5.2 Nuclear Criticality Control by Isetopic Dilution

No such difficulties exist when isotopic dilution 1s used for criticality control. Isotopic
dilution is the addition of “*U sufficient to lower the “*U concentration level below that at which
nuclear criticality can occur. It is important to note that all uranium isotopes have the same
chemical characteristics if they are in the same chemical form; therefore, the 2*U will not separate
from fissile uranium in any normal chemical process during processing or after disposal.

If the ®°U is declared to be waste, isotopic dilution converts the material from a fissile
material for which nuclear criticality is a major safety concemn into another type of very lowly
enriched uranium (LEU) waste for which nuclear criticality is not a significant concern. This
approach simplifies waste management operations for several reasons. First, it allows the use of
existing waste management facilities, such as HLW vitnfication plants, to convert the uranium into
an acceptable chemical form for disposal. Waste management facilities are not typically designed
to be geometrically safe for criticality control, and chemical reactions within such processes may
separate uranium from other elements that are neutron absorbers. Second, this strategy allows
disposal in a geological repository without creating new and difficult nuclear criticality issues.
Finally, isotopic dilution is compatible with expected repository licensing requirements for the
control of nuclear cniticality.

To ensure criticality control of **U by isotopic dilution with Z°U, the 2*U concentration must
be reduced to ~0.5 wt % (Elam et al. 1996). In terms of nuclear criticality safety, this
concentration is equivalent to **U at an enrichment level of ~1.0 wt %. At these concentrations,
nuclear criticality will not occur in a geological environment, over time, nor in waste processing
operations that have not been designed for fissile materials. For a quantity of **U also containing
some >°U, the amount of DU (0.2 wt % *’U) in grams required to ensure criticality control by
isotopic dilution in a water-moderated system 1s given by Eq. 3.1.

The use of isotopic dilution to control nuclear criticality in a disposal facility has already been
chosen as the preferred option for criticality control for disposal of any excess > U that is disposed
of as waste (U.S. DOE 1996a; U.S. DOE 1996b). The strategy discussed herein for 2*U follows
the strategy for excess 2°U. Some of the *U inventory also contains **U sufficient such as to be
classified as HEU;, that is, it contains >20 wt % “°U. Because the technical, regulatory, and
economic factors relevant to HEU are, therefore, relevant for much of the U materials, the use of
a common criticality control strategy for the disposal of any fissile uranium materials is further
supported.

37
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41
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Rose Gottemoller, U.S. Department of Energy, Office of Nonproliferation and National
Security, 1000 Independence Ave., S W., Washington, D.C. 20585.

Peter Gottlieb, TESS, 1180 Town Center Dr., Las Vegas, Nevada 89134

Thomas Gould, Lawrence Livermore National Laboratory, MS-L.186, P.O. Box 808,
Livermore, Cahfornia 94551.

Leonard W. Gray, Lawrence Livermore National Laboratory, MS-1.394, P.O. Box 808,
Livermore, California 94551.

M. Haas, U.S. Department of Energy, Rocky Flats Environmental Technology Site, P.O.
Box 464, Golden, Colorado 80402,

James C. Hall, U.S. Department of Energy, Qak Ridge Operations Office,
20 Administration Rd., Oak Ridge, Tennessee 37831.

Reginald Hall, Advanced Integrated Management Service, Suite B3, 575 Oak Ridge
Turnpike, Oak Ridge, Tennessee 37830. '

Bill Halsey, Lawrence Livermore National Laboratory, MS-L369, P.O. Box 808, Livermore,
California 94551.

. Roger Henry, P. O. Box 1625, Idaho Falis, Idaho 83415-3805.

. M. K. Holland, Westinghouse Savannah River Company, P.O. Box 616, Aiken, South

Carolina 29802.

. J. D. Hulton, U.S. Department of Energy, Office of Fissile Materials Disposition, DOE/MD-

4, Forrestal Bldg., Rm. 66-092, 1000 Independence Ave., S W., Washington, D.C. 20585.

Tim Hunt, U.S. Defense Nuclear Facilities Safety Board, Suite 700, 625 Indiana Ave., N'W |
Washington, D.C. 20004,

Brent Ives, Lawrence Livermore National Laboratory, 7000 East Ave.,
Livermore, California 94550.

N. C. Iyer, Westinghouse Savannah River Company, Savannah River Site, Bldg. 773-A,
Aiken, South Carolina 29808,
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Leslie Jardine, Lawrence Livermore National Laboratory, MS-L186, P.O. Box 808,
Livermore, California 94551.

Bill Jensen, U.S. Department of Energy, MS 1101, 850 Energy Dr_, Idaho Falls, Idaho
38401-1563.

Hoyt Johnson, U.S. Department of Energy, EM-66, Forrestal Bldg., 1000 Independence
Ave., S W_, Washington, D.C. 20385.

Ed Jones, Lawrence Livermore National Laboratory, 7000 East Ave., 1.-634,
Livermore, California 943550,

Professor William E. Kastenberg, University of California-Berkeley, Department of Nuclear
Engmeering, Berkeley, California 94720-1730.

R. Kenley, U.S. Department of Energy, GA-242, EM-66, 1000 Independence Ave., S W |
Washington, D.C. 20585,

J. F. Krupa, Westinghouse Savannah River Company, Savannah River Site, Bldg. 773-41A,
Aiken, South Carolina 29808.

Terry Lash, U.S. Department of Energy, Office of Nuclear Energy Science and Technology,
1000 Independence Ave., S.W., Washington, D.C. 20585.

Rodney Lehman, U.S. Department of Energy, DP-24, 19901 Germantown Rd.,
Germantown, Marvland 20874,

Leroy Lewis, Lockheed Martin Idaho Technologies Company, P.O. Box 1625,
Idaho Falls, Idaho 83415.

Cary Loflin, U.S. Department of Energy, P.0O. Box 5400, Albuquerque, New Mexico 87185-
5400.

Kathy Martin, U.S. Department of Energy, DOE/GC-52, 1000 Independence Ave., S W,
Washington, D.C. 20585.

Herbert Massie, U.S. Defense Nuclear Facilitics Safcty Board, Suite 700, 625 Indiana Ave.,
N.W., Washington, D.C. 20004.

Mal McKibben, Westinghouse Savannah River Company, Savannah River Site, Bldg. 773-
41A, Rm. 123, Aiken, South Carolina 29808.

T. McLaughlin, Los Alamos National Laboratory, ESH-6, P.O. Box 1663, Los Alamos, New
Mexico 87545.

Don McWhorter, Westinghouse Savannah River Company, Savannah River Site, Bldg. 704-
F, Aiken, South Carolina 29808.

Dave Michlewicz, U.S. Department of Energy, DOE/ER-7, 1000 Independence Ave., SW,,
Washington, D.C. 20585.

Lawrence E. Miller, U.S. Department of Energy, DOE/NE-40, 19901 Germantown Rd.,
Germantown, Maryland 20874
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Ed Moore, Westinghouse Savannah River Company, Savannah River Site, Bldg. 773-41A,
Rm. 125, P.O. Box 616, Aiken, South Carolina 29808.

Jim Nail, Lockheed Martin Idaho Technologies Company, P.O. Box 1625,
Idaho Falls, Idaho 83415.

Dave Neiswander, Advanced Integrated Management Services, Inc., Suite B-3, 575 Oak
Ridge Tumpike, Oak Ridge, Tennessee 37830.

Jon Nielsen, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, New Mexico
87545,

Dawvid Nulton, U.S. Department of Energy, Director, 1000 Independence Ave., S W,
Washington, D.C. 20585.

H. B. Peacock, Westinghouse Savannah River Company, Savannah River Site, Bldg. 773-A,
Aiken, South Carolina 29808.

Lee Peddicord, Texas A&M University, Strategic Programs, 120 Zachry, College Station,
Texas 77843-3133.

David Pecler, Westinghouse Savannah River Company, Savannah River Site, Bldg. 773-
43A, Aiken, South Carolina 29808.

. Per F. Peterson, University of California-Berkeley, Department of Nuclear Engineering,

Berkeley, California 94720-1730.

Tish Price, U.S. Department of Energy, Rm. 700N, 1301 Clay St., Oakland, California
94612.

Victor Reis, U.S. Department of Energy, Assistant Secretary, Defense Programs, 1000
Independence Ave., SW., Washington, D.C. 203585.

D. R. Rhoades, U.S. Department of Energy, DP24, 1000 Independence Ave., Washington,
D.C. 20585-0002.

Susan B. Roth, University of California, Nuclear Materials and Stockpile Management
Division, Nuclear Materials Management Office, MS E524, P.O. Box 1663, Los Alamos,
New Mexico 87545.

Greg Rudy, Westinghouse Savannah River Company, Acting Manager, Savannah River
Operations Office, Road 1, Aiken, South Carolina 29802.

149 S. S. Sareen, TRW, Suite 800, 2650 Park Tower Dr., Vienna, Virgmia 22180.

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Glenn T. Seaborg, Senior Technical Review Group, Suite 800, 600 S. Tayler, Mail Box
12098, Amarillo, Texas 79101.

Thomas F. Severvnse, Westinghouse Savannah River Company, Nuclear Matenals
Stabilization Program, Bldg. 704-7, Aiken, South Carolina 29808.

Linda Seward, 1daho National Engineering and Environmental Laboratory, P.O. Box 1625,
Idaho Falls, Idaho 83415.
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184-185.

Robert Stallman, U.S. Department of Energy, 850 Energy Dr., Idaho Falls, Idaho 38401.

Kent Sullivan, Westinghouse Savannah River Company, Savannah River Site, Bldg. 773-
41A, Aiken, South Carolina 29808.

Dean Tousley, U.S. Department of Energy, MD-4, 1000 Independence Ave. S W,
Washington, D.C. 20585.

John Tseng, U.S. Department of Energy, DOE/EM-66, Forrestal Bldg., Rm. GA-242, 1000
Independence Ave., S W., Washington, D.C. 20585,

Bruce Twiming, Manager, Albuquerque Operations Office, Kirtland Air Force Base,
Pennsylvania & H Street, Albuquerque, New Mexico 87116.

Terry S. Vail, Westinghouse Hanford Company, R2-534, P.O. Box 1970, Richland,
Washington 99352.

Richard VanKonynenburg, Lawrence Livermore National Laboratory, MS-L269, P.O. Box
808, Livermore, Califorma 94551.

K. E. Waltzer, U.S. Department of Energy, Savanna River Site, Bldg. 703-F, Aiken, South
Carolina 29808.

John Wilcynski, Manager, Idaho Operations Office, 765 Lindsay Blvd., Idaho Falls, Idaho
83401.

Jeff Wilhams, Idaho National Engineering and Environmental Laboratory, P.O. Box 1625,
Idaho Falls, Idaho 83415.

M. Williams, Louisiana State University, Nuclear Science Center, Baton Rouge, Louisiana
70803.

Wendell L. Williams, U.S. Department of Energy, MD-3, Forrestal Bidg., 6G-081,
1000 Independence Ave., S W, Washington, D.C. 20585.

C. R. Wolfe, Westinghouse Savannah River Company, Savannah River Site, Bldg. 773-A,
Aiken, South Carolina 29808.

Jesse L. Yow, Jr., Lawrence Livermore National Laboratory, University of California, 7000
East Ave., Livermore, California 94550.

Loong Yong, Advanced Integrated Management Service, Suite B3, 575 Oak Ridge
Tumpike, Oak Ridge, Tennessee 37830,

Office of Assistant Manager of Energy Research and Development, P.O. Box 2008, DOE-
ORO, Oak Ridge, Tennessce 37831-6269.

Office of Scientific and Technical Information, P.O. Box 62, Oak Ridge, Tennessee 37831,