ORNL-TM-13553.txt

ORNL/TM-13553

Disposition Options for Uranium-233
C. W. Forsberg, E. C. Beahm, L. R. Dole, A. S. Icenhour, and S. N. Storch

Chemical Technology Division
Oak Ridge National Laboratory”
Oak Ridge, Tennessee 37831-6180
Tel: (423) 574-6783
Fax: (423) 574-9512
E-mail: forsbergcw@ornl.gov

L. C. Lewis
Idaho National Engineering and Environmental Laboratory
Idaho Falls, Idaho 83415
Tel: (208) 526-3295
Fax: (208) 526-4902
E-mail: llewis @inel.gov

E. L. Youngblood

Advanced Integrated Management Services, Inc.
Oak Ridge, Tennessee 37830

June 1, 1999

 

"Managed by Lockheed Martin Energy Research Corp., under contract DE-AC05-960R22464 for the
U.S. Department of Energy.
CONTENTS

LIST OF FIGURES . . . . . e e e s s s s s X
LIST OF TABLES . . . . e e e e e e s s s s X1
ACRONYMS AND ABBREVIATIONS . ... st X111
PREFACE . . . XVil
EXECUTIVE SUMMARY . . . e e e s s s X1X
ABS T R ACT . .. XXX1
1. INTRODUCTION .. . e e e e e s s s 1
1.1  OBIECTIVES . . . e e e 1

1.2  SELECTION OF STORAGE AND DISPOSAL OPTIONS . ............... 1

1.3 CAVEATS .. 1

1.4 SPECIAL ISSUES ASSOCIATED WITH?*U ........... ... . ... 2

1.5 REPORT ORGANIZATION . . ... e e s 2

2. CHARACTERISTICS AND INVENTORIES OF *¥U .. ... ... 3
2.1 CHARACTERISTICS . . . . e e st 3
2.1.1  Chemical . .. ... ... 3

2.1.2 Radiological .......... ... . . . . .. 3

2.1.3  NUClear . . ... 8

2.1.4 Institutional . . .. ... ... 8

2.1.4.1 Safeguards ........... ... ... 8

2.1.4.2 Waste Management . ...............uiuueeinanennnn.. 9

2.2 INVENTORIES OF 28U . . .. s 10
221 INVENTOTIES . . . .o et e e e e e e e e e e e e e e e 10

2.2.1.1 Clean 23U . ... 10

2.2.1.2 CEUSP 2PU ... s 10

2.2.1.3 LWBR 22U ... 16

222 QUuality ... 16
CONTENTS (continued)

3. GOALS AND CONSTRAINTS ... e e
3.1 GOALS .. e
3.2 STORAGE OPTIONS AND CONSTRAINTS . ...... ... .. .. ...
3.2.1 Storage Policy Options . ............ ... ..
3.2.1.1 Store As-Is ..o
3.2.1.2 Isotopically Dilute to Non-Weapons-Usable ***U for
Future Use . . ... ...
3.2.1.3 Isotopically Dilute to Critically Safe **U .. ...............

3.2.2  Impact of Potential Uses of ***U on Storage and Disposition Options . . .

3.2.2.1 Medical Applications . .............. ...
3.2.2.2 Low-Mass Reactors for Deep-Space and Other Special-Purpose
MISSIONS .« . vttt e
3.2.2.3 Analytical Tracer . . ........... .. ..
3.2.2.4 Nuclear Weapons Research ... ........................
3.2.2.5 Reactor Fuel CycleResearch . .........................
3.3 DISPOSAL OPTIONS AND CONSTRAINTS . ... ... ... . . . ...
3.3.1 Criticality Control . ...... ... ... . . . . . .
3.3.2  Waste Definitions and Legal Constraints . .......................
3.3.3 Interactions Between and Among Disposal Site Criticality Control,
Waste Volumes, and COSES . . . . .t o e e e e e e e e e
3.3.4 Economics and Schedules ............. ... ... ... ... .. . ...
3.3.5 Storage for Future Disposal ............... .. ... ... ........
4. TECHNICAL DESCRIPTIONS OF OPTIONS . ... ... ... .. . . . ..
4.1 CONVERT **U TO HLW GLASS WITH DU FROM HLW SLUDGE
(DISPOSAL) . ..
4.1.1 General Description . . .. ....... .. .ttt
4.1.2  SubOPUONS . ..o
4.1.3 Current SRS BaselineOption . . . ................. ... ..........
4. 1.4 ISSUES ..
4.2 CONVERT **U AND DU TO A URANIUM-ALUMINUM ALLOY

(STORAGE AND DISPOSAL) . . .o oot oo e

4.2.1 SNF and >**U: Similar Materials and Similar Goals .. ..............
4.2.2 Melt-Dilute Technology . ............. .. .. .. ... ...

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4.3

4.4

4.5

4.6

CONTENTS (continued)

4.2.3  Uranium-233: Specific Uranium-Aluminum Alloy Issues and Options . .

4.2.3.1 Other Chemical Species . ..............couuiiiiin...
4.2.3.2 Direct Processing of CEUSP ***U Containers ..............
4.2.3.3 Product Volumes ................. ... . ... ...........
4234 Option Variants . . ... .. ...ttt

424 ConcluSiONnsS . . . ..o ittt e

AQUEOUS NITRATE BLENDING OF **U WITH DU (STORAGE AND

DISPOS AL) ..

4.3.1 Characteristics of Aqueous Option . ........... ... ..
4.3.2  Process Description . . .......... ...t

FUSION-MELT BLENDING (STORAGE AND DISPOSAL) .............
4.4.1 Process DesCriptions . .. ... .. ...ttt

4.4.1.1 Glass Fusion Melt Process Description ..................
4.4.1.2 Boron-Oxide FusionMelt ............................

4.4.2  UNCErtaINtIES . . . v o vt et e e e e e e e e e e e e e e e e e
4.4.3  ConcClUSIONS . . . .ottt e e

DRY GRINDING, BLENDING, AND SINTERING (STORAGE AND

DISPOS AL) ..

4.5.1 Process Description . . ....... ... it
4.5.2  Variant Isotopic Mixing Methods with Dry Blending ...............

4.5.2.1 DryBlendingOnly............. ... .. ... .. .. ... .....
4.5.2.2 Dry Blending and Consolidation . ... ....................
4.5.2.3 Dry Blending with High-Temperature Sintering ............

4.5.3  Safety Issues with Dry-Blend Options . .........................

4.5.4 Mechanisms of Isotopic Dilution with High-Temperature Sintering and

the Implications on Process Design . . ...........................
4.5.5 Conclusions . ........ ...

CHEMICAL DILUTION: WASTE THRESHOLD (DISPOSAL) ...........

4.6.1 Definition of Waste Threshold .......... ... .. .. ... .. ... .......
4.6.2 Disposition Options . .. ...... .ttt

4.6.2.1 Convert to Waste: Process with CH TRUW ..............
4.6.2.2 Convert to Exception Case Material: Stand-Alone Disposal . . .

4.6.3 Special Issues with Waste Threshold Options . . ...................

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81

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4.7
4.8

4.9
4.10

4.11

4.12
4.13

4.14

CONTENTS (continued)

CAN-IN-CANISTER (DISPOSAL) ... ... . o .
RH TRUW PROCESSING (DISPOSAL) ....... ... ..

4.8.1 ORNL

4.8.2 Hanford . . . ... ... ..

CH TRUW PROCESSING (DISPOSAL) ....... ... ... ...
LWR FRESH FUEL (DISPOSAL) . ... .

4.10.1 General Description . . . ........ ...ttt
4.10.2 Past EXpPerience . ... ... ... ...
4.10.3 Conversion of Off-Specification HEU and *’U to LWR Fuel .........

4.10.3.1 NeUtroniCs . . . oo v vttt et e e e ettt e e
4.10.3.2 Alpha ACtVILY . ... ... .
4.10.3.3 Gamma ACUVILY . ...ttt e e

4.10.4 ConcluSIONnS . . . ..ot ittt e e e

DEEP BOREHOLE (DISPOSAL) . ... ... ..

4.11.1 General Description . ... ....... ...ttt
4.11.2 Waste Form and Site Characteristics ..................ccouun....
4.11.3 Deep-Borehole Alternative Applications . ........................

4.11.4 Issues

4.11.5 Conclusions . . . . ..ot e e

GREATER CONFINEMENT DISPOSAL (GCD) (DISPOSAL) ............
SPACE (DISPOSAL) . .. vt o et e e e e

4.13.1 General Description . ... ....... ... .t
4.13.2 Waste Form Characteristics . ... ..........iine ...
4.13.3 Space Disposal Options and Destinations ... .....................

4.13.4 Issues

4.13.5 ConclusSions . . . .. ov ittt

SUBSEABED (DISPOSAL) ...... ... . . . ..

4.14.1 Engineering DescCriplion . . ... ......... ittt
4.14.2 Performance Capabilities ............... . ... ... .. ..

4.14.2.1 Predictability . ........ . ... ... . . . ..
4.14.2.2 Intrinsic [solation . . .. .......... .. ... ... ... ... ... ....
4.14.2.3 Independent Isolation Mechanisms ......................

4.14.3 Institutional Considerations . . . . ... ... ..o

4.14.4 Issues

85
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127

128
128
CONTENTS (continued)

4.15 DISPOSAL AS SNF (DISPOSAL) .. ...ovvieenn.,
4.16 SHALLOW-LAND BURIAL (DISPOSAL) . ...............
4.17 ELECTROMETALLURGICAL PROCESSING (STORAGE AND

DISPOSAL) ..o oottt e e

5. CONCLUSIONS ... i

5.1 SPECIAL CHARACTERISTICSOF**U .................
5.2 STORAGE AND PROCESSING OPTIONS ...............
5.3 DISPOSAL OPTIONS .. ... .. .

6. REFERENCES . ... . . e

Appendix A
Appendix B
Appendix C

Appendix D

RADIATION LEVELS FROM**U ...............

RECOVERY OF **Th FROM **U FOR MEDICAL USE ..........

RECOVERY OF URANIUM FROM CEUSP CANS ..

CURRENT TECHNICAL CONSTRAINTS FOR WIPP

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139
139
139
140

143

B-1

C-1

D-1
Fig.
Fig.
Fig.
Fig.
Fig.
Fig.
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Fig.
Fig.
Fig.
Fig.

Fig.
Fig.

Fig.
Fig.
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Fig.
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Fig.
Fig.
Fig.
Fig.
Fig.

Fig.
Fig.
Fig.

Fig.

2.1
2.2
2.3
2.4
2.5
3.1
3.2
3.3
3.4
3.5
3.6

4.1
4.2

4.3
4.4
4.5
4.6
4.7
4.8
4.9
4.10
4.11
4.12

4.13
4.14
4.15

4.16

LIST OF FIGURES

Different fissile materials require different handling procedures .............
Gamma exposure for 1 kg of 2*U with 100 ppmof 22U .. .................
United States inventory of separated *U . .. ........ ... .. ... .........
Uranium-233 inventory is currently stored in different containers . ...........
Chemical composition of the Z*Uinventory . ...........................
Uranium-233 Storage OPtONS . . . . v i et e et e et e e e e ettt
Potential uses for 22U .. ... ... ... ...
Flowsheet for *’Bi production for treatment of cancer ....................
Recoverable “Th from the U.S. **Uinventory .........................

Technical WIPP WAC constraints . . . . . ... oottt e e e e

Impact of criticality control strategy on waste volumes to WIPP-type disposal

ST o e e e e e e e e e

Conversion of surplus ***U and HLW (containing DU) to HLW glass ........

Isotopic dilution of #*U with DU in a uranium-aluminum alloy for long-term

storage or disposal ... ... ...
Uranium-aluminum phase diagram .. ................ ... .. ...........
Aqueous processing option for **U long-term storage or disposal ...........
Isotopic dilution by the aqueous-nitrate-blending process ..................
Borate-fusion melt for ***U long-term storage or disposal .. ................
Glass-fusion melt (alkali glass) option for disposition of U ... ............
Simple dry-blend processing for ***U long-term storage or disposal options . . . .
Self-diffusion of uranium in uraniumoxide .............. ... ... .. .....
Chemical waste threshold disposition option for #°U in a WIPP-type facility . . .

Definition of **U-containing materials . .. .. ..............iuinenon....

Chemical waste threshold ***U disposition option: Coprocessing with

CH TRUW . e e e e e e e e e e e
Can-in-canister U disposition option . . .. ..........uuuriiiieenennn...

Can-in-canister for disposition of >*U or plutonium . .. ...................

Co-processing of liquid RH TRUW and ***U for disposition in a WIPP-type

TEPOSIEOTY . o v vttt e e e e e e e e e e e e

Uranium-233 disposition by co-processing with CH TRUW ... ... .........

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Fig.
Fig.
Fig.

LIST OF FIGURES (continued)

4.17  Conversion of **U to LWR fuel, irradiation, and disposal of the SNF in YM 99
4.18 Borehole disposal of U .. .. ... ... .. ... 108
4.19  Process flow diagrams for the immobilized and direct deep borehole disposition
AltErNALIVES . . . . . e e 110
.4.20  The deep borehole disposal concept for immobilized disposal of coated ceramic
Pellets TN GroUul . . ... oot 111
421 GCDoptions for 2PU .. ... 115
.4.22  Spacedisposal of 27U .. ... 118
.4.23  Subseabed disposal Options . ... ........ .. 124
.4.24  Characteristics of seabed disposaloptions .. ........................... 125
.4.25  Co-disposal of *°U (defined as SNF) with HLW ... ..................... 130
.4.26  Disposition of **U by dilution to LLW (<100 nCi/g) with shallow-land
diSposal . ... 133
.4.27  Disposition of **U by isotopic dilution with DU using the electrometallurgical
PIOCESS - o vt e ettt e e e e e e e e 136
.B.1  Flowsheet for *"’Bi production for treatment of cancer .................... B-4
ig. C.1  CEUSP can assembly for monolithic uranium oxide storage . ............... C-4
Table ES.1
Table 2.1
Table 2.2
Table 2.3
Table 3.1
Table 3.2
Table 4.1
Table 4.2
Table 4.3

Table 4.4
Table 4.5
Table 4.6
Table 4.7
Table 4.8
Table 4.9
Table 4.10

Table 4.11
Table 4.12
Table 4.13
Table 4.14
Table 4.15
Table 4.16

Table 4.17
Table 4.18
Table 4.19
Table 4.20
Table 4.21

LIST OF TABLES

Characteristics for storage or disposal of dispositioned **U . ..............
Characteristics of weapons-usable materials . . .........................
Quality of major batches of separated **U ininventory ..................
Inventory of **U SNF (not part of disposition program) .. ................
Uranium-233 uses, applicable ***U categories, and isotopic requirements . . . . .
Summary of WIPP WACs, as appliedto U .. ........................
Summary: Convert **U to HLW glass with DU from HLW sludge .........
Summary: Uranium-aluminum melt dilute option. . .....................

Specific volume and mass of the final uranium-aluminum alloy for different
products for an initial 1 kg *?U . ......... ... .. ... .

Summary: Aqueous nitrate blending of **UwithDU . ...................
Product volume and mass per kilogram of **U for aqueous processing . . . .. ..
Summary: Convert **U to borate fusionmelt .........................
Summary: Convert **U to glass by fusionmelt .. ......................
Product volume and mass per kilogramof Z*U . ........................
Summary: Dry grinding, blending, and sintering .......................

The time required for the tenfold dilution of atoms, when the mean self-diffusion
distances, <r>, of the uranium atoms UQO, are 2.16 times the original particles'

16 P
Summary: Chemical dilution: waste threshold ........................

Summary: Can-in-canister

Coprocessing *°U with RH TRUW forms ...................
Convert Uto CHTRUW form . .........................

Summary:
Summary:
Summary: Convert **Uto LWR fuel ............. ... ... ..........

Composition of candidate materials for down-blending by mixing with
SRS DU .o

Composition of proposedblends . . ........ ... ... . .. ... .. ... ...
Comparison of proposed blends with ASTM specifications . . ..............
Summary: Characteristics of deep borehole disposal ....................
Summary: GCD of U ... ... .

Summary: Spacedisposal ......... ... .

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42

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50
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56
58
60
64

73
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87
92
95
98

102
103
103
106
113
117
Table 4.22
Table 4.23
Table 4.24
Table 4.25
Table 4.26
Table D.1
Table D.2
Table D.3

Table D.4
Table D.5a

Table D.5h

Table D.6

Table D.7a

Table D.7b

LIST OF TABLES (continued)

Potential forms for nuclear waste in spacedisposal . . . ...................

Summary: Subseabed disposal

Summary: Disposeas SNF .. ... ... ... .
Summary: Shallow-land disposal as LLW .. ..........................
Summary: Electrometallurgical process ... .......... ... .............
Summary of WIPP WAC as appliedto ™*U . ..........................
WIPP CH TRUW container and assembly weight criteria . ................

Maximum number of CH TRUW containers per TRUPACT-II and authorized
packaging configurations . . ........... ... ...

Comparison of uranium isotope and **Pu characteristics .................

Summary of CH TRUW acceptance criteria requirements that pertain to U
waste thresholds

Summary of RH TRUW preliminary acceptance criteria requirements that
pertain to *’U waste thresholds

Maximum allowable fissile material, expressed as ***Pu FGE for CH TRUW in
the TRUPACT-IL . . . . ... e e e e e s i

Comparison of WIPP-based nuclear criteria for various radionuclides
comprising CH TRUW

Comparison of WIPP-based nuclear criteria for various radionuclides
comprising RH TRUW

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131
132
135
D-5
D-7

D-8

D-10

D-11

D-12
ACRONYMS, ABBREVIATIONS, SCIENTIFIC NOTATION, AND UNITS OF MEASURE

Al Aluminum

AMWTF Advanced Mixed Waste Treatment Facility
ASTM American Society for Testing and Materials
BAC Boeing Aerospace Company

Bi Bismuth

Cd Cadmium

CEUSP Consolidated Edison Uranium Solidification Program
CH contact-handled

DNFSB Defense Nuclear Facilities Safety Board
DOE U.S. Department of Energy

DOT U.S. Department of Transportation

DU depleted uranium

DWPF Defense Waste Processing Facility

EIS environmental impact statement

EPA U.S. Environmental Protection Agency
ES&H environmental, safety, and health

FGE fissile gram equivalent

GCD greater confinement disposal

GTCC greater-than-class-C

HEPA high-efficiency particulate air

HEU high enriched uranium

HLW high-level waste

IAEA International Atomic Energy Agency

ICPP Idaho Chemical Processing Plant

INEEL Idaho National Engineering and Environmental Laboratory
L liter

LLNL Lawrence Livermore National Laboratory
LEU low-enriched uranium

LLW low-level waste

Xiii
ACRONYMS, ABBREVIATIONS, SCIENTIFIC NOTATION, AND UNITS OF MEASURE

(continued)
LWBR light-water breeder reactor
LWR light-water reactor
MeV million-electron volts
MOU Memorandum of Understanding
NA not applicable
NASA National Aeronautics and Space Administration
NRC U.S. Nuclear Regulatory Commission
NTS Nevada Test Site
NWTRB Nuclear Waste Technical Review Board
ORNL Oak Ridge National Laboratory
PA performance assessment
ppm part(s) per million
RCRA Resource Conservation and Recovery Act
R&D research and development
RH remote-handled
S&S safeguards and security
SAR safety analysis report
SNF spent nuclear fuel
SRS Savannah River Site
SWB standard waste box
t metric ton
TDOP ten-drum overpack
Tl Thallium
TRU transuranic

TRUPACT-II Transuranic Package Transporter-II

TRUW transuranic waste

TVA Tennessee Valley Authority
U Uranium

WACs waste acceptance criteria

X1V
ACRONYMS, ABBREVIATIONS, SCIENTIFIC NOTATION, AND UNITS OF MEASURE
(continued)

WGP weapons-grade plutonium
WIPP Waste Isolation Pilot Plant
WP waste package

YM Yucca Mountain

YMR Yucca Mountain repository
Y-12 Y-12 Plant (Oak Ridge)
20877 Thallium-208

22U Uranium-232

2y Uranium-233

By Uranium-235

28y Uranium-238

2B Bismuth-213

XV
PREFACE

This report is one of several reports which maps the strategy for the future use and disposition of
uranium-233 (***U) and disposal of wastes containing >*U. Other relevant documents from this and other

programs are listed below with a brief description of the contents.

e  ORNL/TM-13550—Strategy for the Future Use and Disposition of “’U: Overview. This
document is a summary of the path forward for disposition of surplus **U. It includes required
activities, identifies what major programmatic decisions will be required, and describes the
potential disposition options.

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

»  ORNL/TM-13552—Strategy for the Future Use and Disposition of “°U: Technical Information.
This document summarizes scientific information on >*U. This includes production methods,
decay processes, and the material characteristics. The requirements for storage and disposal are
also included.

e ORNL/TM-13524—Isotopic Dilution Requirements for *°U Criticality Safety in Processing and
Disposal Facilities. This document determines and defines how much depleted uranium (DU)
must be mixed with ***U to prevent the potential for nuclear criticality under all expected process
and disposal facility conditions.

e  ORNL/TM-13517—Definition of Weapons Usable “**U. This document determines and defines
how much DU must be mixed with ***U to convert the ***U into a non-weapons-usable material.

*  ORNL/TM-13591—Uranium-233 Waste Definition: Disposition Options, Safeguards,
Criticality Control, and Arms Control. This document defines what 2*U-containing material is
waste and what **U-containing material must be treated as fissile material.

*  ORNL/M-6606—Uranium-233 Storage Alternative Trade Study: Final Report. This document
evaluates alternative long-term **U storage options and identifies the costs for each option.

e ORNL/TM-13600—Technical Handbook of *>U Material Properties, Processing, and Handling
Guidelines. This document is a reference handbook for handling and processing ***U.

*  ORNL/TM-13553—Disposition Options for Uranium-233. This document describes and
characterizes alternative disposition options for **U (this report).

XVil
EXECUTIVE SUMMARY

The U.S. Department of Energy (DOE) Fissile Materials Disposition Program (MD), in support of the
U.S. arms-control and nonproliferation policies, has initiated a program to disposition surplus weapons-
usable fissile material by making it inaccessible and unattractive for use in nuclear weapons. Weapons-
usable fissile materials include plutonium, high-enriched uranium (HEU), and uranium-233 (***U). In
support of this program, Oak Ridge National Laboratory led DOE’s contractor efforts to identify and
characterize options for the long-term storage and disposal of excess ***U. Five storage and 17 disposal
options were identified and are described herein.

Storage, as a type of disposition, refers to methods to isotopically dilute **U with depleted uranium
(DU) to convert the ***U into a non-weapons-usable form suitable for long-term storage or future use. This
1s functionally equivalent to down-blending of weapons-usable HEU to low-enriched uranium (LEU).
Disposal, as a type of disposition, refers to processing the ***U and disposing of the material as a waste in a
manor that makes the ***U inaccessible and unattractive for use in nuclear weapons. Some disposal options

involve the down-blending technologies necessary for storage.

ES.1 INVENTORIES AND CHARACTERISTICS OF U

The characteristics of **U are different from those of HEU or plutonium. Consequently, the best
options for the storage and disposal of *’U may be different from those for HEU and plutonium.
Uranium-233 has the chemical characteristics of uranium, but it is an alpha emitter like plutonium and
thus, as a minimum, must be processed within an alpha containment (e.g., in a glovebox). Uranium-233
can be rendered unsuitable for weapons use by isotopically diluting it in ***U to <12 wt %. It could be
further diluted with ***U to minimize nuclear criticality issues associated with waste disposal.

Uranium-233 has a characteristic that makes most plutonium or HEU disposition options not viable for
*¥U. In the production of ***U, some ***U is produced. The ***U has a decay product, thallium-208 (***T1),
which decays to lead and produces high-energy, 2.6-MeV gamma rays. The concentration of **U
determines the radiation shielding required to protect workers. Uranium-233, which contains very low
levels [<1 part per million (ppm)] of ***U, has correspondingly low levels of gamma radiation. Uranium-
233 with higher concentrations of ~?U (greater than a few ppm) and with associated radioactive decay
products requires heavy radiation shielding and remote-handling (RH) operations to protect workers from

gamma radiation. Most facilities designed to process HEU or plutonium are unsuitable to process *°U.

XiX
From the perspectives of long-term storage or disposal, the existing ***U inventory can be divided into
three major categories: (1) clean, (2) Consolidated Edison Uranium Solidification Program (CEUSP), and
(3) light-water breeder reactor (LWBR) **U. The different characteristics of these materials suggests that
the preferred storage or disposal option for one category may be different from that for another category of
.

e Clean #°U. This uranium has very few chemical impurities and is primarily in the form of oxides.

There are ~340 kg of clean **U. The uranium isotopic composition is **U with variable impurity

levels of ~*U that are measured in parts per million. The radiation levels vary widely depending
upon the ***U content.

e CEUSP*’U. The CEUSP material is a single batch of material with unusual isotopic, chemical,
and packaging characteristics. The CEUSP ***U inventory contains ~ 1,040 kg of uranium in ~400
containers. The uranium isotopic composition is ~10 wt % ***U, 76 wt % **U, and a complex
mixture of other uranium isotopes. It contains high concentrations of **U that create a significant
radiation field. The CEUSP material contains large quantities of cadmium—a Resource
Conservation and Recovery Act metal—which may create some unique institutional issues. The
CEUSP material was solidified in the storage containers into a monolithic mass that is physically
bound to the inside walls of the stainless-steel containers.

« LWBR?’U. LWBR *U refers to a collection of unirradiated nuclear fuel. The LWBR **U
inventory contains ~350 kg of ?UQ, which is combined with ~ 14,000 kg of thorium oxide (ThO,)
in the form of unirradiated high-fired ceramic reactor fuel. The uranium isotopic composition is

*¥U with low levels of *U. The presence of the ThO, makes chemical processing of the **U
difficult.

ES.2 CHARACTERISTICS OF STORAGE OPTIONS

The ***U could be stored for future use. Potential uses include medicine, space reactors, analytical
measurements, development of thorium and nonproliferation fuel cycles, and nuclear weapons. There is
current interest in production of bismuth-213 (*’Bi) from ***U for treatment of certain cancers. Bismuth-
213 is a decay product of *U. No decision has been made on what material to keep and what material to
discard.

For some uses, such as medical applications, the U could be down-blended with ***U to non-
weapons-usable “*U and remain useful. Consequently, options for down-blending ***U for storage have
been investigated and are described herein. No decisions have been made on what fraction of any ***U that

1s kept should be down-blended to non-weapons-usable material.

ES.3 CHARACTERISTICS OF DISPOSAL OPTIONS

Some disposition options would transform the ***U and dispose of it as waste, thus rendering it

inaccessible and unattractive for use in nuclear weapons. In most of the viable options, the ***U is

XX
converted to non-weapons-usable material by isotopic dilution with DU. The material is then disposed of

as a waste. Uranium-233 is an alpha emitter like plutonium; thus, the most appropriate disposal facility to

use 1s a geological repository. There are two types of geological repositories.

Yucca Mountain (YM)-type repository. The proposed YM repository is designed for spent nuclear
fuel (SNF) and solidified high-level waste (HLW). These waste forms generate significant heat
and are highly radioactive. To assure repository performance, any waste form sent to YM must
meet a series of waste acceptance criteria (W ACs) that set stringent requirements on the chemical
and physical characteristics of the waste. Uranium-233 waste forms could be accepted at YM
under several sets of conditions, provided that they are in acceptable chemical and physical forms.

Waste Isolation Pilot Plant (WIPP)-type repository. The WIPP repository is designed for the
disposal of defense transuranic waste (TRUW), which contains small quantities of transuranic
elements and generate little heat. Because of the characteristics of the waste and the design of the
repository, there are few requirements relating to the chemical form of the waste.

The WIPP site would be technically acceptable for disposal of **U wastes, but the WIPP enabling
legislation does not authorize disposal of ***U wastes at this site. However, TRUW containing **U
is allowed. The expected future volumes of wastes requiring a disposal site with capabilities
similar to WIPP significantly exceed the authorized WIPP capacity. The United States must
ultimately expand WIPP, build other disposal sites to manage these wastes, or find other disposal
options. Existing ?*U wastes and other wastes that require a WIPP-type disposal facility, but
which can not be sent to WIPP, are packaged to meet WIPP WACs because these criteria define
not only the requirements for waste acceptance at WIPP but also those for long-term storage,
transportation, and disposal of such wastes at future WIPP-type facilities. For a *°U disposition
option that generates WIPP-type wastes, there are two suboptions: (1) send the material to WIPP
after appropriate changes in laws or regulations or (2) add the material to the existing ***U wastes
in inventory. The volumes of “*U wastes in existing storage facilities are significantly larger than
the potential volumes of wastes from any *°U disposition option; thus, the addition of wastes from
¥V disposition activities would not fundamentally alter future **U waste management operations.
Storage of *’U as a non-weapons-usable waste is significantly less expensive with lower
accompanying risks than storing weapons-usable ***U.

ES.4 STORAGE AND DISPOSAL OPTIONS

The **°U storage and disposal options are shown in Table ES.1. In several cases, there are multiple

variants of specific options. Some transformation processes would render the material suitable for

subsequent storage or disposal. For example, aqueous processing may be used to (1) recover medical

isotopes from ***U, (2) down-blend #*U with ***U to non-weapons-usable material for long-term storage

and future use, or (3) convert the ***U into a form for disposal. Conversion of ?*U into a form for disposal

usually includes down-blending the ***U with ***U to minimize safeguards and repository nuclear criticality

issues. For each disposal option, the type of disposal site is listed.

XX1
Table ES.1. Characteristics for storage or disposal of dispositioned >**U

 

 

 

Technical viability Recovery of medical isotopes”

Options with variants Storage Type of disposal site Clean CEUSP LWBR Clean CEUSP LWBR Demonstrated with uranium
HLW glass No YM Yes Yes Maybe Yes Yes No Yes
Uranium-aluminum alloy

Stand-alone variant Yes WIPP Yes Yes No Yes Yes No Yes

Co-process with SNF No YM Yes Yes No No No No Yes
Aqueous Yes YM/WIPP Yes Yes Yes Yes Yes No Yes
Fusion melt

Borate Yes WIPP Yes Yes Yes Future Future No Yes

Glass No YM Yes Yes No No No No Yes
Grind, blend, and sinter Yes YM/WIPP Yes Yes Yes Future Future No Yes
Chemical dilution No WIPP Yes Yes Yes No No No Yes
Can-in-canister No YM No No Yes No No No Yes
RH TRUW No WIPP Yes Yes Yes Yes Yes No No
Contact-handled (CH) TRUW No WIPP No No Yes No No No No
Light-water reactor (LWR) SNF No YM Yes Yes Yes Yes Yes No Yes
Deep borehole® No Borehole Yes Yes Yes No No No No
Greater confinement disposal No GCD Yes Yes Yes No No No No

(GCDY
Space disposal” No Space Yes Yes Yes No No No No
Subseabed” No Subseabed Yes Yes Yes No No No No
Shallow-land disposal” No Shallow-land Yes Yes Yes No No No Yes
Define as SNF No YM No No Yes No No No No
Electrometallurgical® Yes YM/WIPP Yes Yes Yes ? ? No Yes

 

“Yes implies that the medical isotopes can be recovered during processing. Future implies that medical isotopes could be recovered in the future if the material is in storage.
’Footnoted items have very-large technical and institutional uncertainties.
The technical viability of each option for each type of *°U is indicated. Some options are suitable for
only certain categories of “*U. For example, the uranium-aluminum melt-dilute option (described below)
can be used to process clean and CEUSP U, but not LWBR ***U. In this specific case, the basic
chemical processes of this option do not work well with a feed containing large quantities of thorium oxide.
The LWBR material contains large quantities of thorium oxide.

Table ES.1 indicates whether the option allows the recovery of medically useful isotopes from the ***U.
As indicated in the table, there are no current processes for practical recovery of useful isotopes from the
LWBR material. Several options appear to offer practical recovery of material for production of medical
isotopes from the clean and CEUSP categories of materials.

The last column defines whether there has been any experience in processing uranium with the specific

technology. A brief description of these options follows.
ES.4.1 Conversion to HLW Glass (Disposal)

Uranium-233 could be disposed of by mixing the ***U with existing HLW sludges containing DU and
converting the mixture into HLW glass for disposal at YM. The option is technically applicable to all
categories of “U in inventory. This option has several attractive features. It produces the minimum
amount of waste of any option because it uses DU that is currently in the HLW tanks for isotopic dilution.
The HLLW glass is qualified for disposal at YM. The option would use existing or planned facilities at the
Savannah River Site (SRS) or Hanford. The viability of the option for managing LWBR ***U is unclear
because of several technical factors. It is uncertain whether the SRS tanks have sufficient DU to address
criticality issues if the entire inventory of **U was disposed of. Disposal of all material in the Hanford
waste tanks is technically viable but would require potentially longer time-frames. The option allows a

one-time recovery of medical isotopes. The primary uncertainty is cost.
ES.4.2 Melting ***U and DU to Form a Uranium-Aluminum Alloy (Storage and Disposal)

Uranium-233 could be melted with DU metal and aluminum metal to create a non-weapons-usable,
uranium-metal alloy suitable for storage or disposal. If the uranium is in oxide form, it would be converted
to metal in the process. The process has one unique characteristic: the molten aluminum can dissolve
uranium, aluminum, and stainless-steel storage cans. This capability is particularly useful for processing
CEUSP **U because the uranium is in a monolithic form that would be difficult to remove from the storage
can. The option is applicable to clean ***U and CEUSP **U, but it does not appear to be suitable for

processing LWBR U because the LWBR ThQO, is chemically stable in molten aluminum.

XXl
The process is under development at SRS for treatment of aluminum-clad HEU SNF and has been
demonstrated on a small scale with uranium. If an SNF treatment facility is built using this process, the
same facility (with minor modification) could be used to process *U for disposal at YM. Alternatively, a
custom ***U treatment facility could be built for storage of the product or for disposal of the **U as a waste
in a WIPP-type facility. A dedicated furnace might allow the recovery of medically useful isotopes. The
option has potentially low processing costs. However, the final product volumes—because of the added

aluminum—are significantly larger than those with other options. There are technical uncertainties.
ES.4.3 Aqueous Blending (Storage and Disposal)

In the aqueous-nitrate-blending process, solid **U material would be dissolved in nitric acid to produce
an aqueous, uranyl-nitrate solution, which is mixed with a uranyl-nitrate solution of DU. After mixing, the
solution 1s solidified for storage or disposal. Several final storage or disposal form options are possible
(UQ,, UO,, U;0q, cement, and glass). The process would be applicable to all 2°U in inventory. If the *°U
1s to be disposed of as waste, it could be sent to a YM- or WIPP-type repository. This is the only process
that has been used on an industrial scale, and it is the only fully demonstrated technology. It is a proven
process for recovery of medically useful isotopes. Several DOE facilities, with modifications, could
process all **U. The process might be more expensive than some of the other options; however, it is the

only process for which risks are fully understood and cost estimates can be easily developed.
ES.4.4 Fusion-Melt Blending (Storage and Disposal)

In the fusion-melt blend process, the **°U oxide would be mixed with DU oxide powder and solvent
metal oxide powder and then melted in an induction furnace to produce a glasslike material. The melt
crucible would also serve as the product container. The product can be stored or disposed of in a YM- or
WIPP-type repository or used as a pretreatment step to ready the material for disposal in HLW tanks
(Sect. ES.4.1). The process has been demonstrated with uranium, but there are technical and institutional
uncertainties that could impact costs.

There are two process variants: borate-fusion melt and glass-fusion melt. The difference between the
two options is that with the borate fusion melt, the solvent metal oxide would be B,0O; or Na,B,0-,
whereas, for the glass fusion melt, the solvent metal oxide would be a mixture of alkali metal oxides and
silicon dioxide. In the borate-fusion-melt process, a borate would be chosen as the solvent metal oxide to
minimize processing difficulties and maximize the uranium content of the final product. In the glass
fusion-melt process, the solvent metal oxide (usually containing high concentrations of silica) would be
chosen to create a high-quality, insoluble uranium product. This choice would require added processing.
Both variants have been demonstrated with uranium in the laboratory. The two variants have different

uscs.

XX1V
*  Borate-fusion melt. The borate-fusion melt creates a product suitable for storage or disposal in a
WIPP-type repository. All categories of U can be processed. The process is potentially the
lowest-cost isotopic-dilution option (simple process, low-volume product), but technical
uncertainties could impact costs. The option allows for the future recovery of medical isotopes or
the ***U for other uses. It also might be used in combination with another disposition process (such
as disposal in HLW tanks) to optimize overall economics.

*  Glass-fusion melt. The glass-fusion melt creates a high-quality glass suitable for disposal at a
YM-type repository. Clean and CEUSP ***U can be processed; LWBR **U probably can not be
processed because of the chemical effects of thorium on glass properties. Product volume is
significantly larger than that with the borate fusion melt process. The process would not be used
for storage of **°U for future use because of the difficulty of recovering the uranium or medical
1sotopes from the glass.

ES.4.5 Grind, Blend, and Sinter (Storage and Disposal)

In the grind, blend, and sinter process, ***U oxide would be mixed with DU oxide (both in fine powder
form), consolidated, and sintered at high temperatures to produce a final storage or disposal form. It is
applicable to all ®’U. The process is a simplified variant of the process used to make the thorium- or
uranium-oxide reactor fuels. The *°U, after processing, would be suitable for storage or disposal at a YM-
or WIPP type repository. For ***U sent to storage, medical isotopes could be recovered in the future. The
CEUSP #*U (with cadmium) would be suitable at a WIPP-type repository, but tests would be required to

determine suitability for YM-type repositories due to the cadmium in this specific material.
ES.4.6 Chemical Dilution (Disposal)

Uranium-233 could be disposed of as waste if (1) sufficiently diluted in other materials and (2) certain
other requirements are met. This could be accomplished by mixing **U with large quantities of TRUW or
other materials and limiting the quantity of 2*U per waste package. There are policy questions that must

be resolved. This process has the potential for becoming the low-cost disposal option for the ***U.
ES.4.7 Can-in-Canister (Disposal)

Uranium-233 could be disposed of by mixing it with other oxides, converting the mixture into a
ceramic form, and packaging the mixture in small cans. The cans would be placed inside an empty HLW
canister, and HLW glass would be poured around the cans containing >**U. This option is similar to that
proposed for plutonium disposition. There are institutional advantages for using the same processes
proposed for plutonium disposition or for modifying the plutonium disposition process to permit the

inclusion of selected inventories of >*>U material as a feed stream.

XXV
The **U has higher gamma-radiation levels than does plutonium; however, the LWBR **’U could
probably be co-processed with the plutonium with minor modification of the plutonium immobilization
facility. The thorium in the LWBR ***U could replace some or all of the functions of the uranium in the
final product. The clean and CEUSP **U radiation levels are too high to allow processing in the plutonium
immobilization facility. Stand-alone facilities would be required to process these materials in a can-in-

canister option.
ES.4.8 RH TRUW Processing (Disposal)

DOE has large quantities of RH TRUW stored on the Oak Ridge and Hanford sites. It is planned to
process much of this waste into a form acceptable for disposal at WIPP. Some RH TRUW processes, such
as plasma processes, can co-process ~°U, DU, and RH TRUW into a form acceptable for WIPP while
isotopically blending the ***U and DU. No decisions have been made on what processes to use. Depending

upon future waste management decisions, options may exist to dispose of ***U via this route.
ES.4.9 CH TRUW Processing (Disposal)

The United States is planning to build several facilities to process certain existing inventories of
CH TRUW into a form acceptable for WIPP. In this option, the ***U would be co-processed with CH
TRUW and DU to produce a waste acceptable for disposal at WIPP. The **U would be isotopically
blended with the DU. The process is applicable to ***U inventories with low external radiation
levels—primarily the LWBR ***U. It may not be viable for *?U materials with higher radiation levels, such
as the CEUSP material.

If appropriate CH TRUW facilities are built, this is a low-cost option. However, three constraints
determine the viability of this option.

« The option can be used to process only “’U with low gamma-radiation levels. The LWBR #*U

material may meet this requirement. It is relatively pure **’U and is mixed with 14 t of ThO,.
Most of the remaining inventory could not be processed in such a facility.

e The CH TRUW facility must use a process that isotopically blends U with DU. Initially, it was
proposed to build a facility in Idaho with these characteristics (plasma-arc furnace). The treatment
process has been changed, and the new process no longer assures isotopic blending of **U with
DU. No decisions have been made on the CH TRUW processes for other sites. The viability of
the option depends upon waste management decisions on how to treat CH TRUW.

« The wastes must be acceptable for WIPP. Legal uncertainties are associated with mixing ***U with
CH TRUW.

XXV1
ES.4.10 LWR Fuel (Disposal)

The reactor option would convert the **U into LWR fuel for irradiation in power reactors. This option
would allow the beneficial use of the U and would produce an SNF which would be acceptable for
disposal at YM. The **U would be blended with DU to make a LEU oxide feed for the fuel. This
technology has been demonstrated on a large scale. The Shippingport reactor operated using only U as a
fuel. This option would use a disposition path similar to that used to disposition the HEU—blend the
fissile material to <5 wt % as feed to a nuclear fuel fabrication plant.

This would be one of the more expensive options. The production of ***U nuclear fuels would require
special process facilities for uranium purification, conversion to uranium dioxide, and fabrication of
nuclear fuel. No such facilities exist in the United States (facilities do exist in India). Consequently, it
would require large expenditures to build such facilities. Neither existing U or proposed plutonium fuel
fabrication facilities could process ***U because of the radiation fields of the U.S. *°U inventory. The
aqueous processing option for **U disposition, described earlier, is effectively the first step in a multistep
process to make nuclear fuel, which requires the additional steps of uranium purification, fuel fabrication,
and fuel irradiation. Processes such as the aqueous process would be significantly less expensive than this
option.

There is one exception to the previous conclusions. There is a batch of material that contains 42.6 kg
of uranium consisting of 0.8 kg of very pure ***U and 38.7 kg of **U at the Y-12 plant. The low
concentration of the **U, the lack of impurities in this ***U, and its alloying with **U may allow disposition
of this material by converting this material to LWR nuclear fuel. If this were to occur, the conversion
could be done when processing off-specification HEU in a special production campaign with modifications

to the process to allow processing and irradiation of this material.
ES.4.11 Borehole (Disposal)

Uranium-233 could be converted to an appropriate form and disposed of in boreholes at depths of
several kilometers. The same option was investigated for disposal of plutonium. It is technically viable,
but there are major institutional difficulties in siting such a disposal facility. Because of these difficulties,

the option was not pursued for disposition of plutonium.
ES.4.12 GCD (Disposal)

Uranium-233 could be converted to an appropriate form and disposed of in GCD disposal facilities. A
GCD facility disposes of waste at somewhat greater depths than does a shallow-land disposal
facility—typically 10 to 30 m underground. No such facilities currently exist. Major institutional
difficulties would probably be encountered in siting a new disposal facility. Also, it is unclear whether this

option would meet current environmental requirements.

XXVil
ES.4.13 Space (Disposal)

Uranium-233 could be disposed of by launching it into space or the sun or placing it into high earth or
solar orbit. There are major technical, cost, institutional, and environmental, safety, and health issues
associated with launching rockets with radioactive materials into space. This would not be a viable option

for disposal of the small inventory of U,

ES.4.14 Subseabed (Disposal)

Uranium-233 could be disposed of by placing it under the ocean seabed (1) in the mid-ocean or
(2) under a small oceanic island. The option has received major attention in the waste management
community for disposal of HLW and SNF. It has also been considered as an option to create an
international SNF repository to minimize the potential for nuclear weapons proliferation. It is technically
viable and may have long-term environmental advantages over most other disposal sites but a decade and
large resources would be required to develop the technology. It would be a very expensive stand-alone

option for disposal of the small inventory of **U.
ES.4.15 Shallow-Land Burial (Disposal)

The ***U could be diluted by a factor of ~100,000 to meet the definition of low-level waste and then
disposed of in shallow-land disposal facilities. These are significant legal, technical, and economic

uncertainties associated with this option.
ES.4.16 Define as SNF (Disposal)

The LWBR **U could be defined as SNF and disposed of as SNF. The LWBR U is in the chemical
and mechanical form of nuclear fuel and nuclear fuel components. There are significant legal and technical

uncertainties associated with this option.
ES.4.17 Electrometallurgical Processing (Storage and Disposal)

The ***U could be downblended with DU using the electrometallurgical process that produces a
uranium metal product. There are significant technical uncertainties. It is unclear if the metallic waste

form would be acceptable for certain types of disposal sites.

XXVl
ES.5 CONCLUSIONS

Storage and disposal versions of disposition options have been identified and characterized. Multiple
selection criteria will be used to identify the preferred options. No storage or disposal option ranks high in
all criteria; thus, the future selection of preferred option(s) will depend upon the relative importance of
different criteria. Because the various categories of *U in inventory have significantly different
characteristics, the option preferred for one category may be different from the preferred option(s) for other
categories.

Information about the materials and potential disposition paths is sufficient to permit initial evaluation
and elimination of unqualified options. DOE is using the data in this report and other information to decide
what options should be further considered for ***U storage and disposition. DOE report, Summary Report
of the Screening Process to Determine Reasonable Alternatives for Disposition of Uranium-233 (1999):
DOE/MD-0012 describes this process and the results of the initial screening process. However, significant
amounts of additional data must be compiled on technical performance, programmatic timing, and

comparative economics to permit qualitative evaluation and ranking of the options.

XX1X
ABSTRACT

The U.S. Department of Energy, in support of the U.S. arms-control and nonproliferation policies, has
nitiated a program to disposition surplus weapons-usable fissile materials by making them inaccessible and
unattractive for use in nuclear weapons. Seventeen methods to disposition weapons-usable uranium-233
(***U) were identified and characterized. Some methods isotopically dilute ***U with depleted uranium to
convert it to non-weapons-usable uranium. Other methods make the **U chemically or physically

inaccessible.

XXX1
1. INTRODUCTION

1.1 OBJECTIVES

Since the end of the cold war, the United States has been investigating options to disposition surplus
fissile materials that can be used to build nuclear weapons (Forsberg and Krichinsky January 1998).
Disposition is the process of making weapons-usable fissile materials inaccessible and unattractive for use
in nuclear weapons. Fissile materials include weapons-grade plutonium (WGP), high-enriched uranium-
235 (HEU), and uranium-233 (***U). This report describes the disposition options (long-term storage and
disposal) for >**U. The descriptions include both the technical characteristics of particular options and
potential barriers (including institutional) to implementation of specific disposition options.

Storage, in the context of this report, refers to (1) isotopically down-blending weapons-usable ***U with
U to convert it to nonweapons-usable “*U and (2) placing the material in long-term storage. Future
down-blended ***U could be used for certain nondefense applications or be disposed of as a waste.
Disposal refers to all of the steps necessary to permanently dispose of the Z*U in a way that makes the #*U
naccessible and unattractive for use in nuclear weapons. Many of the same technologies could be used for
storage or disposal. However, the preferred technology(ies) for storage may be different from those for
disposal because the requirements for long-term storage (with potential use of down-blended ***U) are

different from the requirements for ultimate disposal of U in a repository.

1.2 SELECTION OF STORAGE AND DISPOSAL OPTIONS

The U.S. Department of Energy (DOE) is using the data in this report and other information to decide
what options should be further considered for ***U storage and disposition. DOE report, Summary Report
of the Screening Process to Determine Reasonable Alternatives for Disposition of Uranium-233 (1999):

DOE/MD-0012 describes this process and the results of the initial screening process.

1.3 CAVEATS
There are two important caveats.

e No decision has been made to dispose of “>U. The option exists to continue to store some or all
of the material in its current form for possible use.

»  This report makes no recommendations on the preferred disposition option(s). However, it does
describe several options and identifies issues associated with implementing specific options. The
choice of option(s) will depend upon multiple economic, technical, and institutional considerations,
including appropriate National Environmental Policy Act review.
2
1.4 SPECIAL ISSUES ASSOCIATED WITH **U

The handling and disposition of ***U are complex because of the unique characteristics of ~°U and
several historical factors. These are summarized herein and described later in this report.

Most facilities designed to process HEU or WGP can not safely process **U. Uranium-233 has the
chemical properties of uranium, but it is an alpha emitter, like plutonium. Uranium-233 contains
impurities that emit high-energy gamma rays. Consequently, hot-cell facilities with special off-gas systems
are sometimes required.

Multiple unique institutional issues are associated with **U. Usually, regulations, laws, and treaties
are developed as needed. Uranium-233 is a minor fissile material; that is, the quantities of this material are
small compared to those of HEU and WGP. Because *’U was never used in the United States on a large
scale, the institutional structure that exists for HEU and WGP does not exist for >**U. An institutional

structure will have to be created for the disposition of **°U.

1.5 REPORT ORGANIZATION

This report describes storage and disposition options. Section 2 summarizes ***U characteristics and
provides inventory data required to define viable options. Section 3 defines the technical and institutional
constraints that are applicable to all options. Section 4 provides detailed descriptions of the options, while
Sect. 5 presents conclusions. The appendixes provide additional technical detail on specific issues

associated with one or more specific option or options.
2. CHARACTERISTICS AND INVENTORIES OF **U

The characteristics of U are very different from those of HEU or plutonium. Consequently, many of
the storage and disposition options for HEU and plutonium are not applicable to **U. Furthermore, the
*¥U inventory can be divided into three major categories where each category of material has markedly
different characteristics than do other ***U materials in the inventory. Some storage and disposition options
that are applicable to one category of material are not applicable (or may require major modifications to be
applicable) to other categories of >*U materials. These unique characteristics and inventory properties are

summarized herein.

2.1 CHARACTERISTICS

Table 2.1 shows the characteristics of U as compared to those of the other two weapons-usable

materials— WGP and HEU.
2.1.1 Chemical

Uranium-233 is chemically identical to natural and enriched uranium (Bereolos April 1998). The same
chemical processes used for depleted uranium (DU), natural uranium, and HEU are usually applicable to
*¥U. However, as a radionuclide, >°U has a higher specific radioactivity than does DU, natural uranium,
or HEU. Consequently, certain radiation-induced chemical reactions occur faster in uranium containing
significant quantities of **U when compared to reactions involving other uranium isotopes. This
phenomenon is important in certain situations such as long-term storage during which the higher radiation
levels of U require that storage containers and *°U storage forms do not contain organics (plastics etc.)

or water that degrades and forms deleterious gases at higher radiation levels.
2.1.2 Radiological

Unlike HEU, the radiological worker-protection requirements for ultrapure ***U are similar to those for
WGP. The primary hazard from U is alpha radiation, which is also the primary health hazard from
WGP. The alpha activity of *°U is three orders of magnitude higher than that of HEU and about one order
of magnitude less than that of WGP. Consequently, the handling and containment requirements

(gloveboxes etc.) for ultrapure **U are similar to those for WGP (Fig. 2.1).
Table 2.1. Characteristics of weapons-usable materials

 

Fissile material

 

 

Characteristic Plutonium HEU ¥y
Production Neutron bombardment of >*U Separation from natural Neutron bombardment of ***Th
uranium
International Atomic Energy 2 5 2
Agency (IAEA) weapons
Category I quantity (kg)
Isotopic dilution limit for None 20 wt % ~12 wt %*
nonweapons®
Isotopic criticality safety Not applicable 1 wt % 0.66 wt %
limit?
Chemical properties Plutonium Uranium Uranium
Radiation
Alpha (relative to HEU) 10* 1 10°
Gamma Low Low Dependent upon **U impurity
Containment Glovebox Laboratory hood Glovebox/shielded hot cell

 

“The 12 wt % U in **®U is based on a technical study (Forsberg March 1998). However, neither U.S. nor international regulations explicitly
address the required isotopic dilution of **U with ***U to convert ***U to nonweapons-usable **°U.
"Isotopic dilution of ***U with ***U to this limit minimizes the potential for nuclear criticality in disposal facilities.
 

High-Enriched L

       
 

 

 

 

 

 

 

 

 

 

 

Weapons-Grad
Plutonium

 

 

 

 

 

 

 

 

 

ORNLDWG98C-488R

Alpha Activity (Relative) = 1
No Significant Gamma

Alpha Activity (Relative) ~10*
Soft Gamma

- Minimal Shielding

- Leaded Gloves Acceptable

 

 

 

 

   

Uranium-233

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

GloveBox Hot Cell

    
  

Alpha Activity (Relative) ~10°

Hard Gamma (2.6 Mev)

- Time Variable

- Gamma From Decay
Product of 232U Impurity

- Clean 233U Processed in
Glove Box

Fig.2.1.Different fissilematerials requiredifferent handlingprocedures.
6

In the production of **U, some ***U is produced. The concentrations of ***U depend upon the specifics
of the production techniques for 2*U. The ***U has a decay product, thallium-208 (***T1), which decays to
lead and produces a high-energy, 2.6-MeV gamma ray. The concentration of U determines the radiation
shielding required to protect workers. Superior-quality U contains very low levels [~ 1 part per million
(ppm)] of *?U and has correspondingly low levels of gamma radiation. Low-quality *’U with higher
concentrations of ~**U (greater than a few ppm) and associated radioactive decay products requires heavy
radiation shielding and remote-handling (RH) operations to protect workers from gamma radiation
(Appendix A).

The *?U in low-quality ***U also impacts the requirements of off-gas systems for processing these
materials. Uranium-232 decays through several isotopes to the noble gas *’Rn. Radon-220 further decays
to ***Tl—the radionuclide with the 2.6-MeV gamma ray. The **’Rn, as an inert gas, can pass through high-
efficiency particulate air (HEPA) filters and then decay to ***T1. To prevent this from happening in a
process system, the off-gas system must contain multiple traps:

« The first HEPA filter traps **U and solid decay products. Radon-220 from the process and solid
materials on the HEPA filter will pass through this HEPA filter.

» Following the first filter, the system must contain charcoal beds, delay lines (~ 10 min), or other
special equipment to hold the radon in the off-gas system until it decays to a solid material that can
be removed from the off-gas with a second HEPA filter.

e A second HEPA filter catches the solid decay products of *’Rn including **TL.

Typical nuclear-grade off-gas systems designed for HEU or plutonium are not acceptable for 2*U
systems with significant quantities of >**U, this limitation restricts which facilities can process this material.

There is an important radiochemical characteristic of this system. If *°U is chemically purified with
removal of the decay products, the ***U with significant concentrations of **U can be processed and
converted into desired forms in gloveboxes and other enclosures without significant radiation exposure to
workers. It takes time (days to weeks) for the ***U radioactive decay products that emit gamma rays to
build up to high enough concentrations such as to require thick radiation shielding to protect the workers.
Very clean processing systems are required for this type of operation. If Z*U-**U contamination remains
in the system, radiation levels will build up with time and can dominate the radiation field from such
processes. The buildup and decay of **U, ***U, and decay products are shown in Fig. 2.2 for ***U with
high concentrations of »*?U. The first set of peaks are from the buildup and subsequent decrease of the
decay products of **U. The second set of peaks are from the buildup and subsequent decrease of the decay
products of #*U. The curve for gamma-ray generation vs time since purification of the uranium shows

that, for some time after purification, the gamma-radiation doses are low.
Gamma Exposure Rate at 1 ft (R/h)

—_i
N

ORNL DWG 98C-157R3

 

—
o
|

o
|

 

/V (2.6MeV)
232\ w . 220p > - 2087 . .. 208pp
(T1/2 =72years) (Ty,=56s) (Ty,=3min) (Stable)
2331} » - »209R;
('I;/Z= 160,000years) (Stable) —

 

I I | | |
1.0 100 10,000 1,000,000
Time (years)

Fig.2.2.Gammaexposureforlkgof?*U with100 ppm of 23?U.
8

Fig. 2.2. Gamma exposure for 1 kg of >**U with 100 ppm of ***U.
The radiological characteristics of ***U have historically determined what uranium was to be managed

as ZU. If a mixture of uranium contains several isotopes, the mixture is handled as ***U—provided that
the **U is the primary hazard. In practice, this procedure implies that uranium materials containing

somewhat >1 wt % >*°U would be handled as >**U.
2.1.3 Nuclear

The nuclear characteristics of *°U are significantly different from those of WGP or HEU. The
subcritical mass limit of *°U is about 520 g [Oak Ridge National Laboratory (ORNL) February 18, 1994].
This is significantly less than that of ***U (700 g) and slightly greater than that of ***Pu (450 g). However,
in certain partly-moderated systems, the critical mass of **U is less than that of plutonium. Facilities
designed for HEU generally are not suitable for the storing or processing of 2*U. The required isotopic
dilution of ***U to minimize the potential of nuclear criticality (0.66 wt % when isotopically diluted with
pure 2*U or 0.53 wt % when diluted with DU containing 0.2 wt % ***U) is less than that for U (1 wt %)
(Elam November 1997).

2.1.4 Institutional

Although ***U has been investigated for many applications, it has not been used on a large scale in the
United States. This has several implications. United States laws, regulations (including DOE orders), and
standards have been developed as needs were identified. The large large-scale production of WGP, low-
enriched uranium (LEU), and HEU have led to the creation of facilities and institutional structures designed
to specifically address issues associated with these materials. For example, the Waste Isolation Pilot Plant
(WIPP) was designed to dispose of transuranic wastes (TRUWSs) from plutonium processing operations. In
addition to the facilities, legal and regulatory structures for the management of TRUW have been created.
The technical characteristics of the WIPP make it suitable for disposal of U wastes, but current law does
not consider U wastes in the WIPP enabling legislation. Thus, wastes containing only *°U can not
legally go to WIPP. There are many other examples. To dispose of **U materials, a set of institutional
issues that are unique to **U must be addressed. The quantities of ***U are sufficiently small such that it
would be cost-prohibitive to build special facilities and to develop a separate institutional framework for
this material. Many of the institutional structures developed for HEU and WGP can be modified and

extended to **U.
2.1.4.1 Safeguards

As a fissile material, *°U is similar to WGP. The IAEA (IAEA August 1993) defines Category I
quantities of weapons-usable materials as 2 kg of WGP, 2 kg of *°U, and 5 kg of HEU. The Category I
quantity 1s that quantity of material requiring nuclear weapons-type security to prevent theft of the

materials.
9

National and international safeguards requirements [DOE orders, U.S. Nuclear Regulatory
Commission (NRC) regulations, IAEA agreements] for weapons-usable materials have been developed for
HEU and WGP; however, the requirements are not developed fully for disposition of surplus ***U. For
uranium containing ***U, these regulatory requirements recognize that only HEU can be made into nuclear
weapons. Natural uranium, DU, and LEU do not require the safeguards and security (S&S) required of
weapons-usable HEU. For disposition of surplus HEU, the U.S. policy (DOE June 1996a; DOE
July 29, 1996) 1s to blend HEU with DU to make LEU for use in commercial nuclear power plants. It is
universally recognized that this process eliminates the use of this material for nuclear weapons and
eliminates the need for weapons-materials-type security.

For ***U, the IAEA regulations (August 1993) do not recognize that mixing ***U with DU will create a
mixture that is unsuitable for the manufacture of nuclear weapons. It is widely recognized within the
technical community that isotopic dilution with DU will eliminate *’U as a weapons-usable material;
however, all 2*U-bearing materials containing significant quantities of *’U are treated as weapons-usable
material. Historically, there never was any serious consideration of converting ?*U to a non-weapons-
usable material; thus, the required regulatory structure was not established. The technical basis for
converting ***U to non-weapons-usable material by diluting it with ***U is understood, but the regulations
and other institutional agreements are not in place. Because one of the goals of the Fissile Materials
Disposition Program is to reduce the risks from these materials in other countries, institutional agreements
as to the level of isotopic dilution that eliminates the weapons potential of ***U are required (Forsberg et al.
March 1998). The isotopic purity that renders 2*U non-weapons-usable (<12 wt % ***U in **U) is less
than that for HEU (<20 wt % ***U in ***U).

2.1.4.2 Waste Management

In the United States, no facilities currently exist for disposing of wastes containing significant
concentrations of ’U. Wastes would be generated from any processing or handling operations.
Historically, wastes containing “**U have been managed as TRUWs (i.e., managed similarly to wastes
containing plutonium) because the primary hazard—alpha radiation—is identical in both waste types; thus,
the same types of disposal facilities are required. WIPP has been built to dispose of defense TRUW .
However, the enabling legislation does not allow the disposal of wastes containing only **U or wastes from
nondefense activities. DOE has several thousand drums of ***U-containing wastes stored at ORNL, Idaho
National Engineering and Environmental Laboratory (INEEL), and Los Alamos National Laboratory. A

WIPP-type facility would be suitable for **U waste disposal.
10

2.2 INVENTORIES OF #*U
2.2.1 Inventories

From the perspectives of long-term storage and disposition options, the separated inventory can be
divided into three major categories: clean, Consolidated Edison Uranium Solidification Program (CEUSP),
and light-water breeder reactor (LWBR) **U. Because of technical, cost, and other factors, many options
can only manage one or two of these *’U categories. Figure 2.3 shows these categories where the area on
top of the cylinder represents the quantity of U and the volume represents the total volume of material
(***U and other materials). Table 2.2 further breaks down the 3 major categories into subcategories. More
detailed inventory information is available in a companion report (Bereolos June 1998). Uranium-233 in
spent nuclear fuel (SNF) and irradiated targets is not included in these numbers.

The inventory contains ~800 kg ***U in ~ 1,800 kg of uranium in a total of 1,505 packages at multiple
sites. Most of the separated U and most of the packages are located at ORNL in the National Repository
for 2°U. The *°U is typically packaged in double containers with the inner container made of stainless
steel or aluminum. Figure 2.4 shows the primary types of storage containers. Figure 2.5 shows the
chemical compositions of the three categories.

In addition to the separated ***U, there is a significant inventory of ***U in SNF (Table 2.3). This
material is shown here for completeness; but, it is not part of the disposition program. As SNF, it meets

the Spent Fuel Standard.
2.2.1.1 Clean **U

The clean ***U consists of five large batches of material plus many small lots. The uranium isotopic
composition is ***U with variable impurity levels of »?U that are measured in ppm (Table 2.2). The
radiation levels vary widely depending upon the ***U content. This uranium is primarily in the chemical
form of oxides with few chemical impurities. Most of the inventory is stored at ORNL in a variety of

containers.
2.2.1.2 CEUSP 23U

The CEUSP ***U was created from the irradiation of a HEU-thorium fuel in the Indian Point Reactor
Unit I, which is owned by the Consolidation Edison Company. The SNF was reprocessed with the *°U
shipped in the form of a uranium-nitrate aqueous solution to ORNL, where it was solidified for storage.
Because all of this material was stored as a liquid solution in a single tank, this is a single, homogeneous
batch of material. It has several unusual properties.

First, the CEUSP materials is a mixture of ~10 wt % **U, ~76 wt % **>U, and other uranium isotopes.

It is #°U isotopically diluted with HEU.
RRRRRRRRRRRRRRR

 

U (Sm

   

 

Total Mass

 

 

 

 

 

Oxidesplus Impurities
Table 2.2. Quality of major batches of separated >*U in inventory”

 

Uranium isotopic composition

Measures of quality”

 

 

Batch Total U By By »y Total U (kg)/ U (mg)/
no. Site Material, packaging (kg) (kg) (kg) (ppm)° 33U (kg) 33U (kg)
CEUSP

1 ORNL U;O4 monolith in >400 welded stainless 1042.6 796.3 101.1 140 10.3 1400

steel cans (CEUSP material)
LWBR
2 INEEL/Radioactive Waste ~ Unirradiated rods and pellets in 172 drums 35.1 0.00 34.0 21 1.03 21
Management Complex?
3 INEEL/Idaho Chemical Unirradiated LWBR fuel with 14 t natural 323.5 0.00 3174 9 1.02 8
Processing Plant (ICPP)“ thorium
Clean

4 ORNL UO, powder in 140 welded inner 67.4 0.00 61.3 165 1.1 180
aluminum cans

5 ORNL U;0¢ monolith in 27 welded stainless steel 65.2 0.00 60.3 15 1.08 16
cans placed in tin-plate cans

6 ORNL UO, powder in 174 stainless steel screw- 96.6 0.00 91.3 7 1.08 7
top cans

7 ORNL UO, powder in 1743 welded stainless steel 46.2 0.00 45 7 1.03 7
plates

8 Y-12 UO, powder in 5 cans 42.6 38.70 0.8 6 53.2 6

9 Various Variety of material forms, packages, and 81.5 <0.1 79.6 <10 1.04

compositions

 

“These data do not represent the entire inventory. Many small batches are not listed. Estimated total quantity of such batches is <1 kg. In addition, the table excludes
23U in SNF and waste streams. Remediation actions associated with the Molten Salt Reactor Experiment may ultimately recover ~40 kg of additional **U to be added to the

clean inventory.
A low number implies higher quality.
‘Based on the total uranium content.

“The Radiological Waste Management Complex at INEEL.

“The ICPP at INEEL.
’The Oak Ridge Y-12 Plant.
13

ORNLDWG 98C-481R2

 

4
3

8.62 in.

 

 

 

 

   
  
 
  
  

 

- 3 Cans per Canister

- Contain Fuel Assemblies,
Rods, Pellets, etc.

 

 

 

 

 

 

LWBR Fuel Storage Canister

 

 

 

 

 

 

«— 12425 1In.

 

    

™ 50% Void

   
     

24 in. Volume
"~ Stainless Aluminum or —»|
Steel Can Stainless Steel ‘

 

 

Clean

- Multiple Container Types
- Powders, Metal, etc.

 

 

Solid (with Can)
Monolithic U30g

CEUSP Can

- Homogeneous 233U

 

Fig. 2.4.Uranium-233inventory iscurrentlystoredin differentcontainers.
Weight Faction

14

ORNL DWG98C-485R3

2.7%Gd,03

 

 

 

 

 

 

 

 

76.6% |
UsOs |

Average|
2.5%
233UO2

 

 

 

 

 

 

 

 

 

Clean CEUSP LWBR
(Many Batches) (Homogeneous (1-12 wt % 233U0;
Material) in ThO,)

Fig.2.5.Chemicalcomposition ofthe 3*U inventory.
Table 2.3. Inventory of U SNF (not part of disposition program)

 

 

 

Uranium isotopics
Total U 23y 22y
Batch no. Storage site (reactor) (kg) (kg) (ppm)
1 Savannah River Site (SRS) (Dresden) 154 high
2 SRS (Elk River) 14.7 high
3 SRS (Sodium Reactor Experiment) 1.1 high
4 INEEL (Ft. St. Vrain) 308.3 90.1 48.3
5 Colorado (Ft. St. Vrain) 822.5 236.0 53.4
6 INEEL (Peach Bottom I) 206.6 20.5 7.1
7 INEEL (Peach Bottom II) 127.8 25.9 58.6
8 INEEL (LWBR) 523.7 220.
Total 927.4

 
16

Second, the CEUSP **U has a high concentration of ***U. This results in a significant gamma
radiation field near the containers.

Third, the packaging system for the CEUSP uranium is unusual—the ***U oxide is a monolithic block
and 1s physically bound to its stainless-steel container. The oxide can be removed from the package only
by chemical dissolution or an equivalent mechanical means. If the container is cut open, the uranium oxide
can not be removed physically from the package using conventional techniques (i.e., powder pouring) and
must be chemically or mechanically extracted. A special loading procedure was used for the CEUSP
uranium. Each stainless steel package was placed vertically in a high-temperature furnace, and the **U
was added as a concentrated uranyl nitrate solution. In the package, the nitrate decomposed to an oxide.
This created a cast-in-place monolithic ceramic in each storage package.

Last, the CEUSP material also contains cadmium and gadolinium, which were added as neutron
absorbers to prevent nuclear criticality of the material while it was in liquid storage—before it was
solidified into its current form. Under some circumstances, the presence of a Resource Conservation and

Recovery Act (RCRA) metal impacts how the material must be handled or disposed of.
2.2.1.3 LWBR #U

The LWBR program investigated the use of *?U-Th nuclear fuels. When the program was shut down,
unirradiated fresh fuel assemblies, fuel rods, fuel pellets, and other assorted materials were placed in
storage at INEEL. While the material is in several package types, it consists of 1 to 12 wt % **UQ, in
high-fired (1,750°C for 12 hours) ThO,. The average assay is ~2.5 wt % **UO, in ThO,. For the
*¥U-ThO, pellets in fuel rods, the assay varies depending upon the location within the fuel rods. There are
also many pure ThO, pellets in some of the fuel rods. The **U in this batch of material is of a high quality
with a variable, but low, *U content. Most of the material contains <10 ppm **U.

The characteristics of this material have two implications in terms of disposition. First, for any option,
there would be a front-end, mechanical sorting process to separate ~?U-ThO, pellets from packaging
materials, clad, ThO, pellets, and other materials. Second, the chemical form of this feed requires that if
this material is to be disposed of, then either: (1) the disposition option must be able to tolerate large
quantities of ThO, or (2) separation of the **’U from the thorium is required before the disposition of the
.

2.2.2 Quality

The three primary inventory categories can be further divided into eight major batches based on the
characteristics of the 2*U (Table 2.2). The remainder of the inventory is a 9th batch (of miscellaneous
small quantities). The quality of the batches can be measured by two indices. Most batches are almost
isotopically pure **U—except for Batches 1 and 8, both of which contain significant quantities of **°U.
For most applications (except nuclear weapons and power reactors), the high ***U content minimizes the
value of the #*U. The second index of quality is the concentration of ***U in the **U. If the #*U

concentration is high, the near-term radiation levels associated with these batches will be high.
3. GOALS AND CONSTRAINTS

This report describes storage and disposal options for 2*U. However, a set of factors (definition of
weapons-usable “U, nuclear criticality, waste management, etc.) strongly influence both the choice of

preferred options and how any specific option would be implemented. These factors are discussed herein.

3.1 GOALS

The primary goal of the disposition program is to make surplus weapons-usable fissile material
unsuitable for weapons use. Weapons-usable **U can be converted to non-weapons-usable >°U by
isotopically diluting it with DU. Isotopic dilution can be used to down-blend **’U for storage and future
use when the application does not require weapons-usable ***U (Sect. 3). Isotopic dilution is used in most
proposed “**U disposal options. Isotopic dilution is the same strategy (DOE June 1996a;

DOE July 29, 1996) as that chosen to convert HEU to non-weapons-usable LEU.

National and international safeguards requirements (DOE Orders, NRC regulations, [AEA agreements)
for weapons-usable materials have been developed for HEU and WGP; however, the requirements are not
developed fully for the disposition of surplus **U. For uranium containing ~°U, these regulatory
requirements recognize that only HEU can be practically made into nuclear weapons. It is agreed to by
DOE orders, NRC regulations, and IAEA conventions that isotopically diluting HEU to <20 wt % *°U
with DU converts the HEU to non-weapons-usable material.

For ***U, the regulations (IAEA August 1993) do not recognize that mixing ***U with DU will create a
mixture that is unsuitable for manufacturing nuclear weapons. It is recognized within the technical
community that isotopic dilution with DU will eliminate the ***U as a weapons-usable material.
Historically, there never was any consideration to convert >**U to a non-weapons-usable material; thus, the
required regulatory structure was not established. The technical basis for defining non-weapons-usable
*¥U by diluting ***U with >**U has been recently developed (Forsberg et al. March 1998). Actions are
being undertaken to arrive at national and international consensus on the level of isotopic dilution required
to make **U non-weapons-usable. The studies indicate that uranium with a ***U content of <12 wt % in
*8U is non-weapons-usable uranium. This concentration is equivalent to uranium with a >U content
<20 wt % in **U.

For mixtures of *’U, ***U, and ***U, effectively non-weapons-usable uranium is defined by the

following formula:

17
18

Weight of 23U + 0.6 weight of *°U
Weight of total uranium

 

< 012 . 1)

For the studies herein, the previous formula was used to determine the quantities of DU needed to make
*¥U a non-weapons-usable material. One kilogram of *°U requires 7.407 kg of DU containing 0.2 wt %

*¥U to convert *’U to non-weapons-usable uranium.

3.2 STORAGE OPTIONS AND CONSTRAINTS

3.2.1 Storage Policy Options

Some or all of the %*U may be placed into long-term storage. Storage is unlike other disposition
options. All disposal options are identical at a fundamental policy level—material is disposed of.
However, different storage options imply different policies. There are three long-term storage options in
terms of the isotopic concentration of *’U in the stored material (Fig. 3.1). The three isotopic

concentrations imply different policies.
3.2.1.1 Store As-Is

The ***U material can be placed in long-term storage containers in its current isotopic form—weapons-
usable Z*U. As a policy, this is (a) the no-action option or (2) a decision that ***U should be kept for all
possible future uses. It is not disposition. Future use may include commercial and defense applications.
The quality of the ***U slowly improves with time as the **U impurity decays at a half-life of 72 years.

In response to the recommendations of the U.S. Defense Nuclear Facilities Safety Board (DNFSB)
Recommendation 97-1 (DNFSB March 3, 1997), DOE (Pena September 25, 1997) has initiated a program
to ensure safe, long-term storage of ***U and development of a long-term storage standard. Because *°U is

weapons-usable material, it must be stored in a high-security vault.
3.2.1.2 Isotopically Dilute to Non-Weapons-Usable ***U for Future Use

The ***U can be isotopically blended with DU and thus converted to non-weapons-usable “*U. This
implies isotopic dilution by a factor of ~7 with DU to <12 wt % ***U in 2*U. The ***U would remain
useful for many, but not all, nonweapons applications. In particular, the **U could be used for production
of medical isotopes (Hall July 22, 1998; Feinendegen and McClure May 1996). Conversion to non-
weapons-usable “U allows the **U to be stored using industrial security levels rather than those required
for weapons-usable material. This implies significant savings in storage costs. The same storage standards

being developed for high-assay material (Sect. 3.2.1.1) would be expected to apply to this material.
19

ORNLDWG98C-340R

233U

v v

Depleted > Isotopic Dilution DOE
Uranium Process Options DNFSB 97-1
/ / \ \ Activities
Aqueous Fusion Uranium-
Melt Blend Aluminum

\ / Alloy

 

0.66 wt % *°U 12 wt % 23U
in 28y in 38y
! { v
Waste Management Low-Security High-Security Vault
Area Storage Facility Storage (Category I:
(Non-Weapons Usable (CategoryIV: Weapons Usable)

and CriticalityControl Non-Weapons U sable)
by Isotopic Dilution)

Fig.3.1.Uranium-233 storageoptions.
20

3.2.1.3 Isotopically Dilute to Critically Safe >*U

The ***U can be isotopically diluted to convert it to non-weapons-usable “**U and to eliminate the
potential for long-term nuclear criticality. The material would then become a waste that is awaiting
disposal. The required isotopic dilution is <0.66 wt % ***U in ***U to eliminate the potential of nuclear
criticality and essentially renders the mixture unusable for all potential applications. It becomes unusable
because of (1) isotopic impurities or (2) the prohibitive cost of processing the bulk of the material to access
some desirable decay product.

This is a distinct and separate option from use options (keep as weapons-usable material or convert to
non-weapons-usable **U) or the disposal options. The disposal options lead to defined end points. This
option leads to long-term storage with disposal to be addressed at a later date. It is potentially viable
because DOE has a significant inventory of existing *°U wastes (Forsberg, Storch, and Lewis
July 7, 1998) with no defined disposal site. Consequently, adding a few hundred additional drums of waste
of isotopically diluted ***U to this existing inventory of several thousand drums would not significantly
increase the waste management activities or costs. No new management system for a new waste form
would be needed. The option (1) does allow the conversion of weapons-usable ***U into a non-weapons-
usable Z*U-containing waste; (2) reduces storage costs because of a lessened need for security
requirements; and (3) does not create new waste management problems. Storage as an end point is further

described in Sect. 3.3.5.
3.2.2 Impact of Potential Uses of >°U on Storage and Disposal Options

There are several potential uses for **U and its decay products. Figure 3.2 summarizes the larger
potential uses. For many applications, the 2*U may be downblended with ***U to nonweapons-usable
material and still be useful. Consequently, many of the disposition options to downblend ***U described in
this report are potentially useful even if the *°U is kept for future use. Table 3.1 defines uses for U and
whether the specific use requires weapons-usable ***U or nonweapons-usable *°U.

The major potential uses of ***U and the potential implications on ***U storage and disposal are

summarized below.
3.2.2.1 Medical Applications

One potential large-scale use for *°U involves one of its decay products, *'*Bi for cancer treatment.
Specifically of interest is that of antitumor antibodies radiolabled with an alpha emitter (Knapp and
Mirzadeh 1994; Geerlings 1993). In this method, the radioisotopes are attached to antibodies that
specifically target cancer cells; the resulting alpha emissions kill these cells with high efficiency. Initial
clinical trials using *Bi on human patients at the Sloan-Kettering hospital in New York City have been

favorable.
21

ORNLDWG97C-149R5

Medical
(Cancer Treatment)

 

 

   

Low-Mass, Deep-Space
Nuclear Power Source

 

 

 

5

   
   
 
 

Analytical
Measurements
(For safeguards,
safety, etc.)

—

Development of
Nuclear Power
Nonproliferation

Fuel Cycles T

 

 

 

Fig. 3.2. Potential uses for 23%U.
22

Table 3.1. Uranium-233 uses, applicable 2**U categories, and isotopic requirements

 

 

 

Applicable #*U category Acceptable isotopic composition in 2*U
Weapons Nonweapons Critically safe
Use Clean CEUSP LWBR >>12 wt %) (<12 wt %) (<0.66 wt %)
Medical Yes Yes No ’ ’ ’
(cancer treatment)
Low-mass reactor Yes No Yes Yes No No
(deep-space reactor)
Analytical Yes No Yes Yes Yes No
(safeguards, etc.)
Weapons Yes No Yes Yes No No
(test, use)
Nonproliferation fuel cycle Yes No Yes ’ ’ No
[research and development
(R&D)]

 

“For these applications, it is unclear what ***U concentrations in 2*U would be acceptable or preferred.

Recovery of *'*Bi for this application involves a three-step process (Fig. 3.3). First, ***U is dissolved in
acid and *Th and its decay products are separated from the uranium by ion exchange. The resulting
thorium-bearing solution contains essentially no fissile uranium, has no nuclear weapons use, and,
therefore, poses no complications in terms of safeguards or nuclear criticality. Next, *’Ac is separated
from **Th and the other decay products. Because actinium is not a part of the decay chain of **U, this
separation removes the undesirable product **®T1 and its precursors. Finally, a biomedical generator system
may be loaded with *’Ac, from which *’Bi may be “milked”.

After the first recovery step, the remaining uranium in solution is resolidified. There are several
options for this **U.

e The *°U can be saved as a future source of **Th or for other purposes. After several years to

allow for ingrowth of **Th and other decay products, the process can be repeated to recover more
*®Th. If the ?**U is to be placed in storage, there is the option of isotopically diluting it with ***U to
nonweapons-usable **U before the ?*U is resolidified. This would increase the mass of uranium

that must be processed in the future to recover **Th; but it would reduce security requirements. It
is not currently known which option is more economic.
 

23

ORNLDWG98C-7582R

 

Currently Stored **U (T1, =160,0

 
 

 

   

DU
Optional

   

 

 

 

 

 

 

 

 
  
 
  

Recover
29Thby
lonExchange

 

 

 

 

 

 

   
  
 

 

   
 

 

ResolidifyandStorein
Standardized Packages
for Storageor Disposal

  

 

 

 

 

 

 

 

Safeguards

No Safeguards 229ThandOther Decay Products

v v

 

 

 

 

 

 

 

 

X\ Separate
225Ac by

lon Exchange

 

#ThT,, = 7340years
225AcT1/2 =10days

 

 

 

 

 

 

 

 

 

 

 

 

HotCell
Hospital
# 225Ac
Y
“Milk” ?"*Bifrom .

Biomedical Bi >
Generator S
System 23BiT,,, =46min

 

 

 

Fig.3.3.Flowsheet for 2I°Bip roductionfortreatmentofcancer.
24

e The *°U may be disposed of as a waste. Because of the long half-life of ***Th (T, = 7,340 years),
separated “*’Th can be used to produce useful *’Ac and *'’Bi for decades. Depending upon
processing loses of *’Th and growth in *"’Bi demand, there may or may not be a need for
additional **Th from *°U after the initial ***Th recovery.

The clean and CEUSP **U can be used in the near-term for medical applications. The LWBR *°U
contains ~ 14 tons of ThO,. It is not practical to isotopically separate the **’Th from the natural ***Th in the
LWBR material. If it was desired to obtain *Th from the LWBR ***U, the following steps would be
required: (1) separate **U from the thorium, (2) store **U for many years to allow buildup of **Th, and
(3) recover newly created **Th from the **U. The current inventory of **Th is shown in Fig. 3.4. The

largest amount is in the CEUSP *°U. This reflects the fact that this material has been in storage for a long

time. This long-term storage has allowed the buildup of the ***Th decay product.
3.2.2.2 Low-Mass Reactors for Deep-Space and Other Special-Purpose Missions

Because “’U has a lower minimum critical mass than ***U or ***Pu (for neutron flux in the thermal
regime), it may be desirable to use it as a nuclear reactor fuel for deep-space missions, for which a
premium is placed on minimizing mass. For this application, weapons-usable ***U would be used to
minimize the launch weight of spacecraft. A space reactor is first put into earth orbit and then is started.
This procedure avoids the need for massive shielding of the reactor before and during launch operations.

The preferred type of nuclear power source to provide electricity for a deep-space mission depends

upon the energy and power requirements.

* For power production levels up to many kilowatts, the minimum-mass nuclear power source is a
radioisotope generator. The currently preferred radioisotope is ***Pu. Nuclear reactors provide
minimum-mass, steady-state power generation at higher power levels. For steady-state power
levels of a few kilowatts to several megawatts, nuclear power reactors fueled with 2*U may
provide the minimum mass (MacFarlane 1963; Lantz and Mayo 1972). For each fissile material, a
minimum mass of that fissile material is required for a nuclear reactor to operate. This minimum
mass is substantially smaller for **U than for °U. Uranium-233 and plutonium have similar
nuclear characteristics; however, the physical properties of uranium in high-temperature space
reactors are substantially better than those of plutonium, and there may be fewer launch safety
issues. These features may make U the preferred material for such applications.

At higher-power levels, the reactor must have large internal heat-transfer surfaces to transfer heat
from the reactor to the electric generator. The reactor fuel assemblies to obtain the heat transfer
require a significant amount of fissile material. In a large nuclear system, the choice of fissile
material does not significantly impact weight because the amount of fissile material needed for heat
transfer far exceeds the minimum critical mass needed for a reactor.
 

  

\

    
  
   
  

///// _

   

////

CEUSP Inventory
101 kg 233U, 796.3 kg 23

     

 

\\\\ - \\\\\\ 124 g 229Th
(0 g 22°Th) \\\\\\\\\\\\\\ . oy

  

Fig. 3.4. Recoverable > Th from the U.S.2*3U inventory.
26

» The total mission energy requirements also impact the choice of fuel for a space reactor. In
missions with large total energy requirements, there must be significant quantities of fissile
materials in the reactor to provide the energy for a long-term mission. Reactor mass is not
determined by the choice of fissile material. HEU becomes the preferred material.

Uranium-233 may also be used for small nuclear propulsion units to boost spacecraft from earth orbit
to deep space (Ludewig et al. 1989; Hyland 1970). These units have moderate-power levels for short times
(<1 h). The interest in using ***U is that it minimizes weight.

For all these applications, only weapons-usable *°U would be used. This includes the clean and

LWBR #**U. The CEUSP material would not be used since it is isotopically diluted with HEU.
3.2.2.3 Analytical Tracer

The **U isotope is used as a calibration spike in the determination of uranium concentrations and
isotopic compositions in materials containing natural uranium or uranium enriched in **U. This type of
analytical procedure is used as part of many safeguards and production operations. There are also other
analytical applications. While the quantities of material used are very small (typically fractions of a gram),

pure U is desired for such applications.
3.2.2.4 Nuclear Weapons Research

Because U is fissile, it has the potential to be used in nuclear weapons. Some ***U may be kept for

research. By definition, only weapons-usable *U is used for this application.
3.2.2.5 Reactor Fuel Cycle Research

The major historical application for ***U has been for research into new nuclear power reactors and
associated fuel cycles. This is also a potential future application. There are four incentives for considering

a **U-thorium fuel cycle.

» The global resources of thorium are about four times greater than those of uranium. If uranium
becomes scarce, thorium 1s a more abundant fertile material to use in reactors to breed nuclear
fuels.

» In thermal reactors, such as light-water reactors (LWRs), thorium fuel cycles breed more fissile
material (**°U) than reactors fueled with LEU.

e SNF and other wastes from the thorium->*"U fuel cycle, when compared to uranium-plutonium fuel
cycles, contain far smaller quantities of long-lived actinides that are a concern in wastes to be
disposed in geological repositories.

«  Some **U-thorium fuel cycles have significantly lower risks of diversion of weapons-usable
material than conventional uranium-plutonium fuel cycles. In power reactors, the impurity >*U
and its daughter products build up to very high levels with correspondingly high radiation levels
associated with the separated **U.
27

Clean and LWBR **U would be used for this application. Because of the unusual isotopics of the

CEUSP #*U, it would not be used for this potential application.

3.3 DISPOSAL OPTIONS AND CONSTRAINTS

There are several constraints that strongly impact multiple disposition options. These are described

herein.
3.3.1 Ciriticality Control

Nuclear criticality must be avoided in any process or disposal facility. Nuclear criticality can be
prevented by controlling equipment and facility geometry, limiting the quantity of fissile material in the
system, adding neutron absorbers, or isotopically diluting >**U with DU. Nuclear criticality avoidance by
isotopic dilution is the addition of ***U sufficient such as to lower the **°U purity level below that at which
nuclear criticality can occur. Isotopic dilution with DU is a preferred criticality control option for many

¥V disposition options based on the following considerations:

e Process criticality control. For relatively small quantities of ***U, strong economic incentives exist
to use already existing waste management and processing facilities. Existing waste management
facilities are not designed for fissile materials or criticality control. The process equipment
geometry in most cases is not critically safe. Avoiding criticality by limiting the quantity of fissile
material in a process system is usually not economical because the minimum critical mass of pure
*¥U is only 520 g for an optimized aqueous system. This implies that the system at any given time
must contain <520 g of **U. The addition of neutron absorbers can prevent nuclear criticality,
but this action requires full control of the chemistry to avoid the potential for separating the 2*U
from the neutron absorber during the process. Because of these considerations, isotopic dilution is
the preferred method for criticality control in many process options.

*  Repository criticality control. 1t 1s difficult to rely on the geometry or chemical composition alone
within disposal facilities to control criticality over geological time frames. Several mechanisms can
cause changes in waste geometry and chemistry, including groundwater transport of uranium and
mechanical disturbances of the waste. If criticality control is to be ensured for thousands of years
by either geometric or chemical control (including neutron absorbers), system performance must be
predictable for these lengths of time. Such predictions are difficult to make, and they are subject to
substantial uncertainties. No such difficulties exist when isotopic dilution is used for criticality
control. In this context, the U.S. Nuclear Waste Technical Review Board (NWTRB), the
Congressionally mandated review board for the proposed Yucca Mountain (YM) geological
repository, has also recommended considering the use of DU to isotopically dilute fissile uranium
materials to prevent the potential for nuclear criticality in geological repositories containing fissile
material (NWTRB 1996). Finally, a recent NRC report made similar recommendations on the use
of DU for criticality control in various disposal facilities (NRC June 1997).
28

» Legal precedent. The environmental impact statement (EIS) (DOE June 1996a) and record of
decision (DOE July 29, 1996) for the disposition of excess HEU recommended isotopic dilution of
the fissile *°U if any HEU was disposed of as a waste. This dilution eliminates the potential for
nuclear criticality in disposal sites. The previous considerations, as they were applied to U, were
the basis for this decision. This precedent suggests that a similar criticality strategy may be used
for any U that becomes waste.

To ensure control of nuclear criticality for ***U by isotopically diluting **U with **U, the *°U
concentration must be reduced to <0.66 wt % (Elam November 1997). In terms of nuclear criticality
safety, this concentration is equivalent to *’U at an enrichment level of ~ 1.0 wt %—a level which will not
result in nuclear criticality under conditions found in processing or disposal facilities. These uranium
1sotopic concentrations avoid the need to control other parameters to prevent nuclear criticality; that is, the
U can be treated as just another radioactive waste. At these concentrations, nuclear criticality will not
occur either in a geological environment, over time, nor in waste processing operations that have not been
designed for fissile materials.

For mixtures of **U and *°U, the amount of DU (with 0.2 wt % **U) in grams (g) required to ensure

criticality control by isotopic dilution in a water-moderated system is the following:

g DU = 188x g 23U + (EO—_Sl) x g of enricheduranium 2)

where

DU
E

grams of DU (0.2 wt % **U) and
the weight percent of °U, where the grams of enriched uranium = total U - **°U.

In the previous equation, >**U and U may be considered to be ***U—providing the atom ratio of the
(**U + #°U): **U does not exceed 1.0. If the quantity of grams of DU calculated is negative, the uranium
material already contains sufficient **U such as to ensure subcriticality; therefore, no additional DU is
needed. For 1 kg of *’U, 188 kg of DU with 0.2 wt % **U is necessary to minimize criticality concerns.

This results in a ***U enrichment of 0.53 wt %.
3.3.2 Waste Definitions and Legal Constraints

For historical reasons, *U disposal was not included in the enabling legislation for many disposal
sites. Consequently, there are a set of legal issues and constraints in terms of disposal of ***U as a waste
material after it has been converted into a non-weapons-usable form. These can be classified into three

categories.
29

Co-processed wastes. 1f the *°U is mixed with another more hazardous waste, the technical and
legal requirements for that waste will usually control the final disposition. For example, if the ***U
1s mixed with DU, converted into reactor fuel, irradiated in a nuclear reactor, and discharged from
the reactor as SNF, the legal definition and requirements for disposal of the SNF would apply.
The fact that the fuel contains *°U is not important in this context. DOE currently owns SNF
containing ~*U, which is being treated for disposal as SNF—not as a**’U waste form. There are
several other similar options such as blending ***U with high-level waste (HLW) (Sect. 4.1) or
blending ***U with TRUW (Sect. 4.6).

Coprocessing of *?U with a waste does not necessarily require mixing of the ***U with the waste on
an atomic scale. For example, it 1s proposed to dispose of excess plutonium by (1) converting the
plutonium to a high-grade ceramic, (2) placing the ceramic in cans, (3) placing the small cans in
empty HLW canisters, and (4) pouring HLW glass around the cans while filling the HLW canister.
In this example, the plutonium ceramic is mechanically distinct and separate from the HLW glass.
The NRC (Paperiello January 25, 1999) has recently stated that “the staff is not aware of any
existing legal or regulatory provisions that would prevent disposal of immobilized plutonium waste
forms in a high-level waste repository.” This precedent indicates codisposal of ***U with various
HLW or SNF forms is potentially viable. It has not been defined how tightly coupled the fissile
waste form must be to the HLW or SNF to be legally treated as HLW or SNF and meet the legal
requirements for waste to be accepted by such facilities.

Direct disposal in a WIPP-type facility. Several disposition options for **U include isotopic
dilution of **U with DU followed by disposal of the material as a waste to a WIPP-type disposal
facility. WIPP is designed for wastes for which the primary hazard is alpha radiation. The
primary hazard of **U is alpha radiation; thus, WIPP is technically suitable for accepting ***U
wastes. Historically, 2?U wastes have been treated as TRUWs. However, the enabling legislation
for WIPP (U.S. Congress October 30, 1992) defined WIPP for disposal of TRUW—elements with
atomic numbers above 92 (uranium). This does not include **U. If #*U is to be ultimately
disposed of as a *°’U waste in WIPP, changes in the enabling legislation will be required if the
material 1s sent to WIPP.

There are related WIPP-type disposition options with different types of constraints. By law

(U.S. Congress 1996), WIPP is authorized to receive and dispose of 175,600 m’ of TRUW
generated from defense operations. DOE, in the EIS for WIPP (September 1997), has determined
that another 142,000 m’ of wastes may require disposal by the year 2033 in a geologic disposal
facility such as WIPP. Much of this waste does not yet exist in packaged form; it will be generated
as old cold-war facilities are decontaminated and decommissioned. These wastes include added
defense TRUW, other government TRUW generated from nondefense activities, and other wastes
that may require geological disposal. The other wastes include existing wastes containing **U that
are in storage—primarily at ORNL and INEEL (Forsberg, Storch, and Lewis July 7, 1998).
Because of the 142,000 m’ of other wastes that ultimately will require disposal, either (1) a second
WIPP-type facility must be built, (2) WIPP must be expanded with congressional authorization to
accept these other materials, or (3) some new option must be identified. As a consequence, there
are effectively two WIPP-type disposal facility options with different constraints: (1) use the
existing WIPP after changes in the enabling legislation allow the disposal of ***U wastes from
disposition activities or (2) store any future >*’U disposition waste designed for WIPP-type disposal
with existing ***U-containing wastes awaiting a disposal facility.
30

» Direct Disposal in HLW-SNF repository. Uranium-233 may be disposed of as a stand-alone #*U
waste in the proposed HLW-SNF repository. The repository is technically designed to accept such
waste forms. Uranium-233 mixed with DU is chemically similar to LWR SNF—uranium dioxide
in metal cylinders. It is not, however, SNF. It would be classified as greater-than-class-C (GTCC)
low-level waste (LLW) not suitable for shallow land burial if regulated under the rules of the NRC
or special case wastes if regulated under DOE orders or could it be classified as HLW by petition
to the NRC. In any case, because of the unique characteristics of the material, unique regulatory
issues would need to be addressed.

3.3.3 Interactions Between and Among Disposal Site Criticality Control, Waste Volumes, and Costs

For many disposal options, there are two variants: (1) isotopically dilute the *°U to 12 wt % in ***U to
convert it to non-weapons-usable material or (2) further isotopically dilute the *°U to 0.66 wt % ***U in
DU to also eliminate criticality issues in disposal sites. In most cases, the disposal options described in this
report assume that the *°U is isotopically diluted to 0.66 wt % ***U in DU. An examination of the
consequences of sending >*U wastes to a WIPP-type facility (Appendix D) with different criticality control
strategies 1s provided herein to understand this constraint on options.

The WIPP waste acceptance criteria (WACs) define the requirements that must be met for a waste to
be sent to a WIPP-type facility for final disposal. There are three criteria that can limit the amount of **U
that can be placed in a drum:

Weight. Drum weight is limited by disposal site handling equipment in WIPP-type facilities and
transport vehicle weight limits.

*  Nuclear criticality. WIPP-type criticality control strategy dependent upon mass limits of fissile
materials in waste packages (WPs). This limit is 200 g *°U/55-gal drum. However, if the ***U is
isotopically diluted with DU to eliminate criticality issues, there is no ***U mass limit per drum.

» Alpha activity limits. The WACs limit the quantity of alpha materials per container with different
limits for untreated vs treated wastes. This criteria is designed to limit the potential consequences
of certain types of accidents.

Figure 3.5 and Table 3.2 summarizes these criteria for a *°U waste. Uranium-233 wastes may
become, depending upon the option, contact-handled (CH) or RH waste. As is evident from the table,
criticality WACs control the maximum quantities of ***U in a WP if fissile mass limits are used as a
criticality control strategy for 2*U. This limit is 200 g of U per 55-gal drum or 325 g ***U per RH
container. If the ***U is isotopically diluted with DU, this limit disappears.

If criticality control does not control the amount of 2*U in a WP, other WACs would limit the amount
of #**U per 55-gal drum. For CH untreated waste, the alpha activity limit of 1.3 kg **U/55-gal drum is
limiting. For CH treated waste, the container weight limit controls the **U content per drum. The
container limit is 450 kg/55-gal drum. To control nuclear criticality by isotopic dilution, ~188 kg of DU
(0.2 wt % **U) is required per kg of 2’U. If the drum weight limit is 450 kg and one assumes that the
waste form is uranium dioxide, the drum can contain no more than 2.2 kg of *U. For RH wastes, the
weight limits and alpha activity limits are by coincidence almost identical with the slightly more restrictive

alpha activity limits restricting the 2*U per container to 16 kg ***U or 5.3 kg per equivalent 55-gal drum.
31

ORNLDWG98C-473R

TRU Waste Category

 

o Contact Remote Handled
Criteria Handled (RH Container
(55-gal Drum) Accepts 3 Drums)

 

<450 kg <3600 kg

 

Untreated <1.3 kg 233y <16 kg 233y

 

Waste

Treated 233 233

Waste <29kg U <16 kg U

Fissile <200 g?3%U/ <325 g 233U/
Mass Limit Drum RH Container

Isotopic

Dilution No Limit No Limit

Limit

 

Fig. 3.5. Technical WIPP WA Cconstraints.
32

Table 3.2. Summary of WIPP WACs, as applied to ***U

 

Waste category

 

Constraint CH RH

 

Container 55-gal drum RH container
(accepts three 55-gal drums)

Weight limit (gross)/container 450 kg 3600 kg
Criticality control strategy
Mass limit/container 200 g/~°U 325¢g
Isotopic limit/container No limit No limit

Radioactivity limit (control of handling risks)

Untreated waste/container 1.3 kg U 16 kg #*U
Treated waste/container 29 kg **U 16 kg #*U
Radiation limit/(surface) <200 mrem/h >200 mrem/h and
<1000 rem/h

 

If one assumes that a metric ton (1 t) of *’U is to be disposed of, about 5,000 drums are required for
disposal if isotopic dilution is not used as the criticality control strategy. If isotopic dilution with DU is
used for isotopic control, as few as 200 drums of waste are generated (Fig. 3.6). The average transport and
disposal costs for TRUW are estimated at about $8,000/55-gal drum (Appendix D). For WIPP-type
disposal sites, using isotopic dilution as a criticality control strategy may lower disposal costs by tens of
millions of dollars per ton of ***U compared to using mass limits of ***U per drum.

Similar interactions between and among repository WACsS, criticality control strategies, safety, and
costs exist or will exist for other disposal sites. For the proposed YM repository, the WACs are only partly
developed. Some analysis to address these issues has been done (SNF Task Team March 1997). In most
cases, disposal site WACs will strongly impact the characteristics and desirability of alternative disposition

options.
3.3.4 Economics and Schedules

For historical reasons, no large production facilities were built to process “*U. Because of the
radiological characteristics of *’U, facilities to process other fissile materials such as HEU and plutonium
usually can not process 2*U. The total quantity of **U (<2 t) is small when compared to the total HEU
(~1000 t) and to plutonium (~ 100 t) produced in the United States. This creates a series of practical
constraints. The cost for new special-purpose facilities for ***U processing will be high. These factors
strongly support an examination of options that use existing facilities and existing processes. The use of
existing facilities and processes implies schedules and other constraints that are imposed by the current

functions of these facilities.
233 U

33

ORNLDWG 98C-381R

  
     

25 ~5000 Drums/
Ton ***U

    

T Y : E_ HRRT
No Isotopic Dilution
WIPP WAC Constraint: 200 g 233U/55-gal Drum

 

 

 

 

  

s
b

~200 Drums/Ton 23U

Isotopic Dilution (0.66 wt % 2>3Uin 238y;
0.53wt% 233UinDU with0.2 wt % 2%°U)
WIPP WAC Constraint: Drum Gross Weight, (>33U, DU)

Fig. 3.6. Impactof criticalitycontrolstrategy onwaste volumestoWIPP-type disposalsite.
34
3.3.5 Storage for Future Disposal

As discussed in Sect. 3.2.1.3, the ***U can be isotopically diluted to convert it to non-weapons-usable
*¥U and to eliminate the potential for long-term nuclear criticality. This option can lead to long-term
storage with disposal to be addressed at a later date. It is potentially viable because DOE has a significant
inventory of existing ***U wastes (Forsberg, Storch, and Lewis July 7, 1998). Consequently, adding a few
hundred additional drums of waste of isotopically diluted ***U to this existing inventory of several thousand
drums would not significantly increase the existing or future waste management activities or costs. No new
management system for a new waste form would be needed. The option (1) does allow the conversion of
weapons-usable “U into a non-weapons-usable **U-containing waste; (2) reduces storage costs because of
a lessened need for security requirements; and (3) does not create new waste management problems.

However, there are several requirements for such a waste.

»  The waste form must meet WIPP WACs. As discussed earlier, the United States does not currently
have a disposal site that can accept >°U wastes. The WIPP is designed for the disposal of alpha-
hazardous wastes. Transuranic (TRU) and **U wastes are both alpha wastes; however, the
enabling legislation does not allow the disposal of *’U wastes in the WIPP (TRUW containing ***U
is allowed into WIPP). While the WIPP WAC:s strictly apply only to wastes accepted for disposal
at WIPP, the WIPP WAC:s define the acceptable nuclear, chemical, and radiological
characteristics for alpha wastes in storage, transport, or disposal. As such, the same requirements
are expected to be applicable to any future waste management facility (storage or disposal) that
accepts similar wastes. For this reason, existing ***U wastes are managed to meet WIPP WACs.

If weapons-usable***U is to be converted into a waste, it should meet the equivalent of WIPP
WAC:s to avoid creating a new unique waste form that would require development of special
storage facilities and, in the long-term, development of special disposal facilities.

o The *U should be isotopically diluted to eliminate criticality concerns. The WIPP facility,
unlike the YM facility, is designed for radioactive wastes containing very small quantities of
plutonium and other fissile materials. The WIPP criticality control strategy is based on limiting the
mass of fissile material per WP to small quantities. This criticality control strategy minimizes
requirements on the chemical or radiological characteristics of the waste; but, it implies large
numbers of waste containers if significant quantities of fissile materials are to be disposed of.
For U, waste volumes can be reduced by one to two orders of magnitude by isotopic dilution of
the **U with **U and using isotopic dilution rather than mass limits as a criticality control
strategy. This implies large long-term cost savings (Sect. 3.3.3 and Appendix D).
4. TECHNICAL DESCRIPTIONS OF OPTIONS

Technical descriptions of each option are included herein. With each description is a one-page

summary of the option.

4.1 CONVERT **U TO HLW GLASS WITH DU FROM HLW SLUDGE (DISPOSAL)
4.1.1 General Description

The **U can be mixed with existing HLW sludges (containing DU) at either the SRS or Hanford and
converted with the HLW sludge to HLW glass (Table 4.1). Specific approaches maybe somewhat different
at the two sites. In all cases, an HLW glass product for disposal at YM 1is produced that meets all
disposition criteria. A generic flowsheet is shown in Figure 4.1 and described later.

Many of the HLW sludges contain large quantities of DU. With appropriate selection of HLW
sludges, the *U will be isotopically diluted with DU currently in the HLW sludges to meet both safeguards
(<12 wt % ***U) and repository nuclear criticality criteria (0.66 wt % *°U when diluted with ***U or
0.53 wt % *°U when diluted with DU containing 0.2 wt % **U). The *°U is also made unrecoverable by
mixing it with HLW sludges. Of all the options, this option is expected to produce the minimum waste
volumes. Minimum waste production is possible because the DU required for isotopic dilution is currently
in the HLW tanks. The tanks at SRS contain about 160,000 kg of U, most of which is DU. The tanks at
Hanford contain about 1.4 x 10° kg of U, most of which is DU. The DU in the HLW sludge at SRS is
sufficient such as to dispose of the CEUSP ***U and perhaps the entire ~*U inventory. Some of the tank-
farm DU is committed for other criticality control purposes and additional analysis is require before it can
be determined if all ***U could be accepted. The DU in the HLW sludge at Hanford is sufficient such as to
dispose of several times the existing inventory of **°U.

The addition of ***U to the HLW sludge is not expected to (1) alter processing of the HLW sludge,

(2) change significantly the characteristics of the final HLW glass, or (3) impact acceptance of the HLW
glass by the repository. The HLW sludge already contains large quantities of uranium. For example, the
SRS HLW glass is expected to average ~2 wt % uranium because of the uranium currently in the HLW
tanks. The additional **U is small in mass compared to the uranium currently in the sludge and thus is not
expected to alter sludge or glass chemistry. Some ***U from early site activities is already in the tanks.
The HLW is highly radioactive; thus, the added radioactivity from the ***U with ***U impurities is not

significant.

35
36

Table 4.1. Summary: Convert >*U to HLW glass with DU from HLW sludge

 

Application
Acceptable **U feeds

Method to convert to non-weapons-usable
233
U

Disposal site
Criticality control strategy

Technical description

Assessment

Advantages

Disadvantages

Evaluation

Disposal
Clean, CEUSP, LWBR?

Isotopic dilution/radiation

YM-type repository
Isotopic dilution

Excess ~°U is dissolved in acid, converted to an acceptable
chemical form to be sent to HLW tanks, the ***U feed is
mixed in the tanks with HLW sludge containing DU, the
mixed ***U and sludge is converted to HLW glass, and the
HLW glass is sent to the Yucca Mountain Repository
(YMR). The process uses existing facilities at SRS or
proposed facilities at Hanford. There are alternative
pretreatment options such as fusion melt (Sect. 4.4) to
prepare the *°U for transfer to the HLW tanks.

Clear path to final repository disposal with a high-quality
waste form.

Allows recovery of medical isotopes before disposal.

Capable of dissolving ***U from CEUSP storage cans.

Much of the DU in the SRS tank farm has been committed
for criticality control of other fissionable materials. The
available DU in the tank farm can dispose of most of the
material; but, it is unclear if all U can be disposed of.
Adding DU would significantly impact ***U disposal costs
at this site (Hanford HLW sludge could accept the total
inventory, but old facilities would have to be restarted or
new facilities built to dissolve the **U).

There are uncertainties associated with LWBR ***U.

This is an attractive disposition option—particularly for
the CEUSP ***U. The CEUSP **U is the most radioactive
U in the inventory and contains cadmium. These
characteristics are not issues if this disposition option is
used. The primary uncertainty is cost.

 
37

ORNL DWG 98C-430R
Process Facility

 

233U

 

 

 

 

 

 

 

 

 
 

 

 

 

 

 

 

 

 

 

Sodium
Nitric Acid Hydroxide
Neutron —3 0 47
Absorber * Yy Y V/ Slurry
Acid
Solution
3 U Dissolver Feed Adjustment for
Tank Farm Acceptance

 

 

HLW Tank Farm

 

 

 

HLW Tank w ith

Depleted-Uranium HLW

YM-Type Repository

 

—

 

 

 

 

 

 

 

 

 

 

 

/
/

Molten
< Glass ;
HLW W
Glass GlassMelter

Fig.4.1.Conversion of surplus 23U and HLW (containingDU)toHL Wglass.
38

Converting *°U to glass also ensures that all chemical repository waste form criteria are met.
Specifically, the CEUSP ***U contains cadmium—an RCRA material. Existing HLW also contain these
materials. The HLW glass is designed to accept these materials and produce a chemically nonhazardous
glass acceptable to the repository. This implies that removal of cadmium and other neutron absorbers from
*¥U is not required if it is to be converted to HLW glass.

The option does allow the recovery of “**Th for medical applications before disposal of the #*U. This
variant requires adding an ion-exchange column for **Th recovery between (a) the dissolution step and

(b) adjusting the ***U feed for acceptance by the HLW tanks farm.
4.1.2 Suboptions
Within this option, there are a series of suboptions that can impact economics.

»  Waste minimization. To minimize ***U processing time and, hence, *’U operational costs, it is
desirable to add the **’U to the minimum amount of HLW. To minimize HLW glass production,
the ***U should be mixed with sufficient HLW sludge such that no clean DU or other neutron
absorber for criticality control needs to be added to the waste. Dilution of ***U with added DU or
other neutron absorbers is to be minimized because added DU or other neutron absorbers generate
added HLW glass logs at a cost of $0.5-2 x 10%log (processing and disposal costs). There is a
trade-off between these two strategies to minimize costs. For example, at SRS there are three
HLW tanks (26, 33, and 34) with 55,000 kg of DU. The remaining DU is spread over a larger
number of tanks. Adding **’U to a small number of tanks is substantially simpler and cheaper than
spreading the ***U among many tanks with each tank containing only small quantities of U and
DU.

e Mixing options. There are multiple engineering options for addition of the **°U to the HLW
sludge. The mixing can occur in the main HLW feed tank, in the tank farm, or in the smaller
process feed tanks in the vitrification facilities. For some equipment options, the ***U may need to
be partially diluted with other neutron absorbers or DU to assure that nuclear criticality does not
occur during process operations.

o Feed preparation options. The ***U to be sent to the HLW tanks could be processed into a form
that meets waste tank acceptance criteria using (1) existing facilities such as the canyons near the
HLW tanks or (2) prepared elsewhere with direct transfer from a shipping cask to the tank farm.
Aqueous processing is the baseline method to prepare feeds to the HLW tanks. However, other
processes such as the fusion melt process (Sect. 4.4) may also be used to prepare the feed.

This option, like all co-processing options, does impose schedule constraints on the disposition of ***U.
The **U must be in the tanks when the HLW sludge containing the DU is to be processed. There is a

window of time when the option exists at SRS and a second window of time when the option exists at

Hanford.
39
4.1.3 Current SRS Baseline Option

A baseline option (Forsberg October 31, 1997) has been defined for disposition of ***U by mixing with
HLW sludge and conversion of the sludge into HLW glass at SRS. The sludge would be converted into
HLW glass at the Defense Waste Processing Facility (DWPF)—a facility currently converting HLW
sludge to HLW glass. At the Hanford site, a program to convert HLW sludge and liquids into HLW glass
is underway; however, it has not progressed to the point where a baseline option can be defined. The

defined SRS option consists of the following steps.

Dissolution. The **’U is dissolved into nitric acid. Neutron absorbers (gadolinium nitrate, DU
nitrate, etc.) are added, as required, to the solution to eliminate concerns about nuclear criticality
during all operations through conversion of the material to HLW glass. The neutron-absorber
mixture is chosen to minimize final glass volumes. Much of the existing Z*U inventory contains
some neutron absorbers (Sect. 2).

*  Chemical adjustment. The acid solution is neutralized with sodium hydroxide, and the chemistry
1s adjusted to meet the WAC for the SRS HLLW tank farm. The change in pH creates a slurry with
any neutron absorbers and the uranium in the solid phase. The slurry has chemical and nuclear
properties such that it can be sent to any tank in the tank farm using existing HLW transfer lines
and equipment.

» Transfer to HLW tank. The slurry is sent to the HLW tanks with high concentrations of DU.

e Sludge mixing. The HLW sludge is recovered from the tanks containing the ***U and the DU. The
recovery process creates two streams: a sludge and a liquid. The uranium (**U and DU) remains
with the sludge. The liquid is separately processed into a LLW stream and a concentrated HLW
stream that 1s mixed with the HLW sludge and sent to a feed tank and then the DWPF for
conversion to HLW glass.

To produce high-quality glass, a homogeneous HLW stream must be fed to the DWPF. Each
specific HLW glass composition must be qualified to meet waste form acceptance requirements of
the repository. Qualifying an HLW glass is a complex operation. The HLW liquids and sludges at
SRS are not uniform and vary from tank to tank. To create a uniform, qualified HLW glass, HLW
sludges and liquids from up to six HLW tanks may be intermixed to produce a single micro-batch
of uniform feed. This batch of feed allows the DWPF to operate up to 3 years producing a single
type of qualified HLW glass.

Complex mixing operations and chemical feed adjustments assure a uniform, quality feed for the
DWPF. This mixing also assures complete mixing of the ***U with the DU with elimination of the
*¥U as a weapons-usable material and minimization of long-term criticality issues. A
homogeneous, uniform mixture of uranium isotopes in the HLW glass is assured by the
requirements to make acceptable HLW glass.

» Vitrification. The HLW feed is sent to the DWPF and converted to HLW glass for the geological
repository.
40

For the previous case, it is estimated that 5 L of added HLW glass will be produced per kilogram of
*¥U. About 1.9 L of glass are produced per kilogram U, assuming that the glass density is 2.5 g/cm’, the
waste loading in the glass is 25 wt %, and the chemical form of the uranium in the glass is equivalent to
uranium trioxide (UQO;). In addition, there will be other neutron absorbers and process chemicals. The
quantities of these materials are not well defined.

SRS is currently planning to dispose of limited quantities of ?*U in its HLW tanks. This establishes
some precedent for disposition of larger quantities of **U in the HLW tanks. The SRS has ~50 kg of
irradiated targets containing ~50 g of **U. It was originally planned to process these targets to recover the
*¥U. When a surplus of U developed, the targets were placed in storage. The targets have degraded
with time and were identified in the DNFSB Recommendation 94-1 (May 26, 1994) as a potential safety
concern. To address this safety issue, DOE has decided (DOE October 1995; DOE December 12, 1995;
and DOE February 8, 1996) to dissolve the targets in nitric acid in 2003, add sodium hydroxide to the
solution to create a slurry, and pump the slurry to the tank farm for eventual conversion to HLW glass.
Detailed planing for this activity has not yet started.

The basic process for irradiated targets is identical to that proposed for disposition of ***U, but there
are differences. The smaller quantity of material allows the addition of large quantities of DU or other
neutron absorbers to the dissolver to control nuclear criticality without creating ultimately large additional
quantities of HLW glass. For larger quantities of ***U, the economics of HLW glass vitrification create
strong incentives to minimize additional HLW glass production. The second difference is that the existing
dissolvers in the SRS canyons can be used (without modification) to dissolve the irradiated targets. The
SRS facilities were designed and operated to dissolve such irradiated targets. The separated ***U inventory
1s in a variety of containers—many of which can not be directly fed to the existing SRS equipment. In this
context, processing the separated ***U in inventory will either require modification of one of the existing
dissolvers or installation of a small special dissolver in one of the hot cells associated with one of the

canyon facilities. These issues are further discussed in Appendix C.
4.1.4 Issues
A number of issues have been identified.

* Schedule. The HLW option exists only for a limited window of time at SRS and another window
of time at Hanford. The U must be disposed of when the HLW with the DU is being processed
into glass. This may happen within a decade for SRS, but will occur further into the future at
Hanford. There is a second schedule issue at SRS. The site is considering shutdown of its canyon
processing facilities. Both processing canyons may be closed in the foreseeable future. These
facilities can be used for the required acid dissolution step. While special-purpose facilities can be
built, there are economic incentives to use existing facilities. The windows of opportunity are
currently being defined.
41

* Available DU in HLW sludge. The quantities of DU at SRS are limited. The potential exists for
other programs to commit to the use of this resource in disposal of other waste streams. The
quantities of DU in sludge at Hanford are very large and, thus, do not appear to be a constraint
under any scenario.

» Technical. There are several technical issues that must be addressed. It is not currently believed
that any of these are a constraint, but some work is required to confirm this. For the SRS option
defined previously, it must be confirmed that the gadolinium and ***U will not separate in the
chemical operations that occur in the tank farm and DWPF. Criticality control measures must be
defined. This impacts final HLW glass volumes. An analysis of dissolver options is also required
(Appendix C).

4.2 CONVERT **U AND DU TO A URANIUM-ALUMINUM ALLOY (STORAGE AND
DISPOSAL)

4.2.1 SNF and ?**U: Similar Materials and Similar Goals

Aluminum-based SNF is being consolidated at SRS for treatment, packaging, interim storage, and
preparation for ultimate disposal. This SNF was discharged from domestic research reactors, foreign
research reactors, and SRS production reactors. Much of this SNF contains HEU (SRS April 1997).
From a technical perspective, the ***U, which is in stainless steel or aluminum cans, has similarities to the
SNF: (1) the primary component is uranium, (2) a clad (can) surrounds the uranium, and (3) some
radiation is associated with the material.

Most of the HEU SNF is highly radioactive; thus, there are limited concerns about the potential use of
such materials in weapons. However, some of the HEU SNF has low levels of radioactivity, which results
in concerns about the potential for recovery of HEU that could be used in weapons. Consequently, DOE
(June 1996Db) is investigating a melt-dilute process that simultaneously melts aluminum-clad SNF and DU
metal to produce a uranium-aluminum alloy suitable for disposal as a waste. Sufficient DU is to be added
such as to convert the HEU to LEU and address repository criticality problems. The **U disposition
program has the similar goals: convert excess weapons-usable “*U to nonweapons-usable material and
convert any material considered waste into a form acceptable for disposal. The same technology is

potentially applicable for 2**U disposition. Table 4.2 summarizes this option. Figure 4.2 shows the option.
42

Table 4.2. Summary: Uranium-aluminum melt dilute option®

 

Application
Acceptable **U feeds

Method to convert to non-weapons-usable
233
U

Disposal site
Criticality control strategy

Technical description

Assessment

Advantages

Disadvantages

Evaluation

Storage, disposal
CEUSP, clean 23U

Isotopic dilution

YM- and WIPP-type repository
Isotopic dilution

Excess 2**U, aluminum, and DU metal are melted in a
high-temperature furnace to create a uranium-aluminum
alloy. The alloy is the final storage or waste form.

Potentially a low cost process to convert *°U to
nonweapons material.

The process can dissolve the U and the packages, thus,
avoid unpackaging operations. This is particularly
desirable for the CEUSP material.

WIPP variant may allow future recovery of medical
isotopes.

Process can not treat LWBR ***U.

The final product volume is significantly larger than with
many other process options. This implies increased
storage or disposal costs.

Viability depends upon results of ongoing research.

Costs are uncertain

This is an attractive option if DOE decides to build a
facility to treat aluminum HEU SNF. The same facility
with little or no modification may be usable for U
disposition. For CEUSP *°U, it is uncertain whether the
material could go to YM. No leachability tests have been
done on uranium-cadmium-aluminum waste forms.

It is a potentially competitive stand-alone option to convert
U into an acceptable storage form or for WIPP-type
disposal.

 

“The process 1s being developed for treatment of aluminum-clad SNF for disposal. Uranium-233 could
be processed using the SNF facility (if built) or processed in a stand-alone facility.
43

 

 

ORMLDWGEREC-374R2
Melt Metal andConvert Oxides to Metal
2830, orotherOxide
(Optional Storage DU (Metal)

 

Can Dissolution) | | Al (Metal)

® Melt MixturetoCreate

Crucible \‘ Uranium-Aluminumalloy

& ChemicalConversion
(33%U0, + 281—»
AlO 5+ 3 Z3UMetal)
Induction ® Optional Dissolutionof Al or
Furnace —— Stainless-Steel StorageCan

® Solidify Ingot inFurnace

Transter Ingot t©o Waste Package and Weld Lid

/’_\qfi Slag (AlO4/Other)
AlLO /Other)

U-Al Metal Alloy

 

   
  

Storage Disposal

 

YM-typeRepository
iCodisposal with
Uranium-AluminumAll oy
from SNFProcessing

WIPP-type
Geological

Reposito
Existing 3% Waste P v

Storage Area

Fig. 4.2.Isotopic dilutionof 231U with DU in a uranium-aluminum alloy forlong-term storage or
disposal.
44

4.2.2 Melt-Dilute Technology

The melt-dilute technology is currently under development at SRS (April 1997) for processing
aluminum-clad SNF. The draft SRS Spent Nuclear Fuel Management Environmental Impact Statement
(DOE December 1998c¢) identifies it as the preferred management option for this SNF. The technology,
without modification, can be used to isotopically dilute ***U for long-term storage or prepare a waste form
for disposal. Initially, the incoming SNF or **U is characterized to determine its uranium content. This
data would be used to determine the amount of DU needed to dilute the HEU or **°U to make the material
non-weapons-usable and to address nuclear criticality issues.

The SNF element or ~*U would be placed in a crucible and melted. DU (in the form of a metal) and
aluminum would be added to the crucible. The final product would be 9 to 67 wt % uranium in the
uranium-aluminum alloy.

If the uranium is in the oxide form, it reacts with the aluminum metal to form a uranium-aluminum
alloy and a slag of aluminum oxide floating on top of the melt. For example, if the feed contains U;Oy, the

following chemical reaction is expected to occur (Adams, Peacock, and Rhode September 9, 1998):

52Al + 3U,0, - 8ALO, + 9UAI, . 3)

This chemistry allows for the processing of aluminum-clad SNF containing oxides and **’U in oxide form.
However, studies have also shown that the chemical kinetics of oxide conversion in aluminum are slower
than for metallic uranium. By addition of elements such as calcium, the oxide is readily reduced and
dissolution becomes rapid. Almost all the **’U in inventory is in the form of oxides.

Three final alloy compositions are being studied at SRS for the uranium-aluminum alloy: 13.2, 30,
and 67 wt % uranium. The final densities of these compositions are about 3.02 g/cm’ (measured),

3.64 g/cm’ (estimated), and 6.36 g/cm’ (estimated), respectively. The uranium-aluminum phase diagram is
shown in Fig. 4.3. The 13.2 wt % composition is the eutectic for the mixture, which results in the lowest
melting point (about 600°C). This composition is currently the preferred option for aluminum-based SNF
melting; however, additional studies are required before a final determination can be made. The other two
mixtures have higher melting points: about 1000°C for the 30 wt % mixture and about 1500°C for the

67 wt % mixture.

The final alloy composition will represent a trade-off between the operating temperature to melt the
alloy, the final solidified volume of the alloy, and performance of the alloy as a waste form. As DU is
added, the uranium composition will increase, thereby increasing the required temperature to maintain the
liquid form (see Fig. 4.3). If SNF is being processed, increased temperatures will drive more fission
products into the off-gas system during processing. Hence, the operating temperature will dictate the
design of the off-gas system. On the other hand, if aluminum is added to maintain the composition nearer
to the eutectic mixture, then the required temperatures will be lower. However, these lower temperatures
will be offset by a higher final volume of the alloy. The ongoing studies of the three proposed compositions

will provide insight into the optimum final composition.
Temperature {*C)

1700

1600

1500

1400

1300

1200

1100

1000

900

800

700

600
500

ORNL DWG 97C-491R2

Aluminum (at.%)

 

 

 

 

 

 

 

 

 

 

 

 

020405060 70 80 85 %0 95 100

N L | |

o] I I I I I I
o L g "\ —
/7 \\\
_ - _
/ N
y < _
/ 1350° C N

| _

,’ 1105’ C

T (¥U) < < —

, = =

2 _
| UosAL{ 6) _
& 758 %BU) 731°C

0L ) eas°c 84'1\\ i °

1 645°C -1 e60.452°C
(o) | 1 641°C 86.8  —
—P
l ; (L e YoeAlda) z g (Al)

0 10 20 30 40 50 60 7 8 90 100
U Al

Aluminum (wt %)

Fig.4.3.Uranium-aluminumphasediagram.

%
46

Fig 4.3. Uranium-aluminum phase diagram.
When the desired final composition of the melted uranium-aluminum alloy has been reached, then the

molten alloy can either be solidified in the crucible or poured into a mold. The solidified form can then be
sealed and placed in dry storage. Canisters containing the diluted alloy will then eventually be shipped to a
geologic repository for disposition.

The alternative uranium-aluminum alloy product compositions are presented in Table 4.3. Two cases
are shown for the dilution of the **U with DU. For the first case, it is assumed that the material is
isotopically blended down to nonweapons material. For 1 kg *°U, 7.407 kg of 0.2 wt % DU would be
required. In the second case, it is assumed that the material is isotopically blended down to eliminate long-
term nuclear criticality concerns. For 1 kg *°U, 188 kg of 0.2 wt % DU would be required. (Note that in

both cases, the amount of DU required accounts for the amount of **U in the DU.)

Table 4.3. Specific volume and mass of the final uranium-aluminum alloy for different products
for an initial 1 kg ***U

 

 

 

Specific volume of alloy Mass of alloy
(L/kg ™) (kg/kg *U)
Composition of alloy Critically Critically
(Wt % uranium) Nonweapons® safe’ Nonweapons* safe’
13.2 21 474 64 1,432
30 8 173 28 630
67 2 44 13 282

 

“Dilution of the **U to a nonweapons enrichment (~ 12 wt % U in ***U).
’Dilution of the ***U to eliminate criticality concerns (~0.53 wt % ***U in DU with 0.2 wt % **U).

The #*°U that is already contained in aluminum packages should be acceptable for co-processing with
the aluminum-based SNF. The remainder of the ***U which is packaged in steel/stainless steel containers
could be repackaged into aluminum containers. Alternatively, it may be possible to process these materials
directly without repackaging because stainless steel is soluble in aluminum alloy mixtures. Further study

of the chemistry may allow such an option.
47

Initial testing of the melt-dilute technology is currently being performed at SRS to determine the
viability of the process (Barlow September 9, 1998; Peacock, Adams, and Iyer September 9, 1998; Adams,
Peacock, and Rhode September 9, 1998; Krupa September 9, 1998). Further tests will be conducted to
determine if the performance of the final waste form is acceptable. An economic (DOE December 1998a),
nonproliferation impact assessments (DOE December 1998b), and a draft EIS (DOE December 1998c¢)
have been recently completed on the process and the program. It is expected that by mid 1999, DOE will

decide whether to proceed to engineering design of the melt-dilute technology.
4.2.3 Uranium-233: Specific Uranium-Aluminum Alloy Issues and Options
4.2.3.1 Other Chemical Species

This option is technically suitable for clean and CEUSP ***U. The option is not viable for the LWBR
*¥U. In the process, uranium oxides are converted to metal by a chemical reaction with aluminum metal.
The LWBR material contains ~350 kg of 2*UQ, in about 14 t of ThO,. Unfortunately, aluminum will not
convert ThQO, to thorium metal with subsequent dissolution of the thorium metal in the aluminum metal.
The ThO, is thermodynamically more stable than Al,O;. Unless the thorium is converted to metal, the
aluminum can not reach the *UQ, inside the ThO, matrix to convert it to metal and dissolve it into
aluminum.

Consideration is being given to recover ***Th from the clean and CEUSP **U for medical purposes.
The same process chemistry that prohibits use of this process on LWBR #*U may allow recovery of ***Th
from the clean and CEUSP ***U inventories when the ***U is being isotopically diluted with DU. In these
two feeds, the thorium content is measured in parts per million “Th—not natural thorium ***Th. Because
of these low thorium concentrations, the uranium-aluminum process will work. However, the thorium will
remain as an oxide. As the *’U oxides are converted to uranium metal producing aluminum oxide, it
would be expected that the thorium oxide would be incorporated into the aluminum oxide to produce a slag
on top of the metal alloy. The thorium can be recovered from this slag. No process experiments have been
done, so the efficiency of recovery is unknown.

The CEUSP **U contains significant quantities of gadolinium oxide (Gd,O,) and cadmium oxide that
were added to assure criticality control. The cadmium oxide is expected to be reduced to metal (Chellew
and Bennett 1961). Some fraction of the cadmium will volatilize into the off-gas system. The fraction of
the cadmium to the off-gas system is strongly dependent upon the chosen operating temperature. The
Gd,O; may remain as an oxide (Chellew, Bennett, and Trice 1961) and would be expected to float to the

top of the melt and form a slag with other oxide materials.
48
4.2.3.2 Direct Processing of CEUSP **°U Containers

The packaging of the CEUSP **U presents significant problems. The **°U is in the form of an oxide
monolith that is integral with the container; i.e., if the container is cut open, the uranium oxides do not flow
out of the container but rather remain attached to the walls. For most options, the uranium oxide will have
to be mechanically removed from the storage cans—a complex mechanical process. The aqueous
(Sect. 4.3) and borate fusion melt (Sect. 4.4) processes can dissolve the uranium oxides out of the storage
cans. The uranium-aluminum alloy process creates the unique option of dissolving the stainless steel can
and the uranium within the can. This simplifies front-end processing.

Studies at SRS indicate that both 304 and 316 stainless steel dissolve in molten aluminum with the
dissolution rate increasing rapidly with temperature. Carbon steel—a potential crucible material— does
not dissolve into molten aluminum. Development tests would be required to determine if a practical

process could be developed to dissolve the ?**U and its storage can.
4.2.3.3 Product Volumes

There is one well-understood disadvantage to the uranium-aluminum melt options. The product
volumes are greater than for most other process options. The aluminum results in a lower uranium density
final product. The volumes can be reduced by producing a higher uranium content alloy but this does
require higher process temperatures with the expected complications. This volume factor is an issue for
¥V disposition but it is not an issue for SNF disposition. Aluminum-clad HEU SNF has a large number
of cooling channels. If the SNF is melted into a monolithic ingot with added DU, the final volume is still

considerably smaller than the original SNF.
4.2.3.4 Option Variants

There are several uranium-aluminum alloy storage and disposition variants. The success of the SNF
program will determine which options are potentially viable for ***U disposition. There are three technical
challenges for the SNF program: (1) develop a workable uranium-aluminum melt process, (2) develop an
off-gas system for the fission products from melting SNF, and (3) qualify the uranium-aluminum alloy for
disposal at YM. If the R&D is successful, a decision can then be made on whether to build a facility. The
*¥U variants are:

»  Use SNF facility. 1f the SNF R&D program is successful and a decision is made to build an SNF

process facility; then, the same facility may be used to process ***U and dispose of the waste at
YM.
49

There is an additional uncertainty associated with CEUSP ***U if the final product is sent to YM.
The proposed YM repository does not accept RCRA (chemically hazardous) wastes. There is a
U.S. Environmental Protection Agency (EPA) test that determines whether a particular waste is
chemically hazardous (i.e., an RCRA waste). No such testing has been done with uranium-
aluminum-cadmium waste forms.

e Build *U process facility and dispose of waste at YM. 1f the R&D program is fully successful,
the option exists to build a separate facility for U disposition. This may be a practical option
because a *’U uranium-aluminum process in an existing facility would cost a small fraction of
what an SNF treatment facility would cost. SNF treatment with this process implies a costly
facility because melting SNF releases volatile fission products to the off-gas system. This, in turn,
requires development of a complex and an expensive off-gas system. This is the major facility
cost. With #**U, there does not exist a high-volume, expensive-to-treat, corrosive off-gas.

e Build **U process facility and send the product to storage or dispose of the material as a waste
at a WIPP-type facility. 1f the final waste form is not fully qualified for YM, the option is to send
the waste to a WIPP-type facility. The existing data indicate that the uranium-aluminum alloy
would meet WIPP WAC. The presence of cadmium in the CEUSP ***U would not be an issue
since RCRA materials are accepted in WIPP-type facilities.

4.2.4 Conclusions

The SRS R&D program has demonstrated the basic uranium-aluminum melt process. The initial
experiments have been successful. Significant work remains to develop an acceptable off-gas system for
SNF (Howell and Zino September 9, 1998; Hodges and Hyder September 9, 1998). Difficulties have been
encountered in qualifying the waste form for YM (Duguid et al. September 10, 1998; Louthan, Wiersma,
and Mickalonia September 10, 1998; Lam, Sindelar, and Peacock September 10, 1998).

If the technology were successfully developed, it would be applicable for the isotopic dilution of **U
with DU for storage or disposition of excess ~*U. There is one exception, the option can not process
LWBR ***U. There are several variants. The viable variants depend upon the success of the SNF R&D

program and SNF programmatic decisions.

4.3 AQUEOUS BLENDING OF **U WITH DU (STORAGE AND DISPOSAL)

4.3.1 Characteristics of Aqueous Option

The aqueous nitrate blending process can be used to convert **U into nonweapons material for long-
term storage or into a waste form meeting all technical requirements for disposal at a YM- or WIPP-type
facility. The option is summarized in Table 4.4 and shown in Fig. 4.4. It is the most versatile U process
and is the historical process used to purify and produce various **U compounds. It allows the option of
recovery of thorium isotopes for medical purposes. It also allows separation of thorium or cadmium
compounds from the feed to produce desired storage or disposal forms for **U-containing materials.

Multiple storage or waste forms can be produced.
50

Table 4.4. Summary: Aqueous nitrate blending of >*U with DU

 

Application
Acceptable **U feeds

Method to convert to non-weapons-usable
233
U

Disposal site
Criticality control strategy

Technical description

Assessment

Advantages

Disadvantages

Evaluation

Long-term storage and disposal
All

Isotopic dilution

YM-type repository, WIPP-type repository
Isotopic dilution

Excess “*U is codissolved with DU in nitric acid and
converted to the desired product or waste. The process can
produce multiple storage and waste forms.

This is the only demonstrated technology.

The process can produce different storage and waste
forms.

The process allows recovery of medical isotopes and
removal of troublesome impurities.

The process 1s potentially more expensive than other
options.

No existing facility is currently capable of this type of
processing. Facility modifications would be required.

This is the historical, standard, industrial process for ***U.
No existing facility is set-up to process existing inventory.
Facilities at ORNL, SRS, or INEEL could be modified to
do this type of processing.

 
51

ORNLDWG98C-383R3

Dissolver
233UO2

Nitric Acid (HNO,) —¥¢ DUO,(NO,),

Separations (Optional)

 

 

 

Th Recovery
- 229Th (medical)

    

P-u-tl‘-i|f1ig:l;‘tl\i’anR) Ex (1?1';' nge Solvent
- Cd Removal Separation Extraction

  

- Decay Products

Thermal Denitrification/ Liquid <€
Consolidation
3N03—+

Compact Sinter Cement In-Can Calcination Vitrification
U;04 U0, U30g/Cement U30g Glass

 

 

 

 

 

     

'&

 
    

YM-type WIPP-type
Repository Geological
Repository

 
 

 

Fig. 4.4.Aqueous processingoptionfor 233Ulong-term storageordisposal.
52

In the aqueous nitrate blending process (ORNL July 13, 1995b), solid ***U material is dissolved in
nitric acid to produce an aqueous uranyl nitrate solution, which is mixed with a uranyl nitrate solution of
DU. After mixing, the solution is denitrated (by heating) to form an oxide powder from which the U
cannot be separated from the ***U without isotopic enrichment. The powder may be either pressed into
pellets or incorporated into grout to provide an acceptable form for transport and disposal. Alternatively,
the solution may also be denitrated in the storage can to produce an oxide monolith similar to the CEUSP
*¥U monoliths. If desired, thorium, cadmium, and other impurities in the **U feed can be removed by
adding a purification step after acid dissolution. A schematic flowsheet for the process is shown in
Fig. 4.5.

4.3.2 Process Description

Uranium-233 material is lifted from the storage well into a shielded cask, which is then transported to
the loading port at the manipulator hot cell. Processing the material is carried out on a batch basis to
control criticality and inventory. After inspection and inventory, the containment vessels are opened, and
the contents are removed. Equipment is provided for cutting the cans open. The material can be removed
by mechanical means or by direct acid dissolution of the material in the can (Appendix B).

Granular or powdered material containing 1 kg **U is converted to a uranyl nitrate solution by reaction
of about 3 L of 4 M nitric acid per kilogram of U (as oxide) to produce ~330 g U/L of solution. A uranyl
nitrate solution of DU of the same concentration is prepared by the reaction of DU oxide (U;0y) with nitric
acid outside the hot-cell area. Dissolution of the thorium oxide-***U oxide fuel currently stored at INEEL
will require the addition of hydrofluoric acid and aluminum nitrate to the nitric acid to aid in dissolution.
After acid dissolution, the option exists to remove useful medical isotopes, cadmium, thorium, or other
impurities from the feed. The standard processes for these separations are ion exchange and solvent
extraction.

Blending is done by mixing measured amounts of ~°U nitrate solution and depleted uranyl nitrate
solution to produce controlled isotopic concentrations of uranium. To produce a free-flowing powder
during denitration, 2 mol of ammonium nitrate per mole of uranium is mixed with the uranyl nitrate
solution during the blending step (the modified direct denitration process).

A one-step conversion of the blended uranyl nitrate solution to uranium trioxide (UQ;) is accomplished
by thermal denitration in a rotary kiln. The process produces a free-flowing fluffy powder. The powder
has a low density (about 1 g/cm’) and is in a form that might be easily dispersed. Several methods are
available to consolidate the material to a safer monolithic storage form (Table 4.5). The powder can be
compacted to 4.3 g/cm’ (about 60% of the theoretical density of UO;) to produce a nondusty, monolithic
form. If desired, the compacted material can be sintered to form higher-density uranium dioxide like the
dry-powder blend process. Alternatively, the oxide may be incorporated into a cement grout at a 50 wt %

loading.
53

ORNLDWG97C-36R3

233y(Input 1k g %)

 

 

 

 

  
 
 

 

 
  
   
  
 

 

 

 

 

 
  
 

 

 

  
  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Depleted Can Removal and
Uranium Size Reduction
l r Nitric Acid Nitric Acid
Uranyl Nitrate Aqueous
Preparation Dissolution
2330, (NO;3), Solution
Blending
Depleted L
UO> (NO3)>
Solution Kiln In-Can
Denitration Denitration
v Y UO3 Powder U308
Grouting | Pellet Pressing Monolith
+Grout +
Packaging I Packaging Packaging
Product: « UO3 in Cement Grout Product: * UO; in Pellets Product: # U;0 3 Monolith
«20.2kg/kg >Ufor *10.1kg/kg 233U *9.9 kg/kg 23U
Nonweapons forNonweapons for Nonweapons
« 454 kg/kg 233U for »227kg/kg 233U for » 223 kg/kg 233U for
Criticality Control Criticality Control CriticalityControl

Fig.4.5.Isotopicdilutionby the aqueous-nitrate-blendingprocess.
54

Table 4.5. Product volume and mass per kilogram of **U for aqueous processing

 

 

 

 

Volume Mass
(L/kg **U) (kg/kg **U)

Processing option <12wt % **U  <0.67* wt % **U <12wt % **U  <0.67* wt % **U
Pellets (UO;) 2.3 52.8 10.1 227
Sintered pellets (UO,) 0.96 21.7 9.5 214
Grout 4.8 108.2 20.2 454
In-can denitration (U,Oy) 3.1 67.5 9.9 223

 

“The required isotopic dilution for criticality control is 0.66 wt % *°U in pure **U. If DU with 0.2 wt % *°U is
used, the final concentration of U is 0.53 wt %. Some of the Z*U must be used for criticality control of the **U in
the DU.

Other process options exist to solidify the liquid uranyl nitrate solution. An immobilized oxide can be
produced by denitrification of the blended uranyl nitrate solution in the product can (similar to that done for
the CEUSP material and described in Appendix C). Using this method, ammonium nitrate would not be
added to the nitrate solution. The solution is slowly dripped into the product can that is in a vertical tube
furnace at a temperature of ~800°C. At these conditions, the water is driven off, and the resultant uranium
nitrate is decomposed to produce a U;04 monolith product. This solidification process may be simpler to
implement on a small scale. The liquid solution may also be fed to a glass melter to produce a uranium
glass. This is a variant of the HLW glass disposition option. The products from any of the immobilization
methods are packaged in doubly contained vessels with welded seals for transport and disposal.

All of the uranium oxides would be expected to be acceptable for long-term storage. There are some
technical uncertainties associated with the grout because of the possibility of generation of hydrogen from
*¥U alpha radiolysis. All of the uranium oxides would be acceptable for disposal in a WIPP-type facility.
All of the uranium oxides would likely be acceptable for disposal in a YM-type facility; however, high-
fired uranium dioxide would probably be preferred. The proposed YMR is designed to accept
63,000 metric tons of 1nitial heavy metal SNF in the form of uranium dioxide pellets in metal tubes. There
will be fewer issues associated with a waste form that is chemically identical to the primary waste form

expected at the repository.
55

The aqueous process can produce several alternative final products with different volumes and masses.
Isotopic dilution of 1 kg of ?**U to non-weapons-grade uranium by the aqueous-nitrate-blending process
will produce 10.1 kg of UO; (8.407 kg U) as pellets having a volume of 2.3 L. Grouting the oxide will
produce 20.2 kg of material having a volume of 4.8 L. In-can denitration will produce 9.9 kg of U,Oq
product having a volume of 3.1 L. Blending 1 kg of **’U to eliminate long-term criticality will produce
227 kg of UQO; pellets (189 kg U) having a volume of 52.8 L. Incorporation of the oxide into grout will
produce 454 kg of grout with a volume of 108.2 L/kg of *U. Denitrification of the blended nitrate
solution in the product can will produce 223 kg of U;0q4 having a volume of 67.4 L. The actual waste
volume per kilogram of ***U will be higher in many cases because other materials present in some of the
23U feeds (thorium oxide, cadmium oxides, gadolinium oxides, etc.). Actual volumes may vary depending

upon the specific process conditions chosen.

4.4 FUSION-MELT BLENDING (STORAGE AND DISPOSAL)

Uranium-233 and DU can be converted to glass by dissolution of uranium oxides into molten glass
formers in a high temperature furnace. The product is a non-weapons-usable glass suitable for long-term
storage or disposal at a YM- or WIPP-type facility. Glass production is a trade-off between processability
(furnace temperature, dissolution time, etc.), glass durability, and high waste loadings. There are two
process variants; one chosen to minimize costs (Table 4.6; Fig. 4.6) and the other chosen to produce a high
quality waste glass that would meet YM-type waste form requirements (Table 4.7; Fig. 4.7).

The first variant is a borate fusion melt wherein the glass former [boron oxide (B,O;) or borax
(Na, B,0O,)] is chosen to minimize processing costs and maximize uranium loadings in the final product. It
can be used to isotopically dilute ***U oxides with DU oxides and create a solid borate fusion melt suitable
for long-term storage or disposal at a WIPP-type facility. The product would not be acceptable for a YM-
type facility unless the borate was removed in a subsequent processing step. The process could be used to
dilute *°U to 12 wt % **’U in ***U (nonweapons ***U) that is stored for potential future use. The product
characteristics would allow relatively simple recovery of uranium or decay products such as ***Th (for
medical applications) at a later date. The process may also be used as a pretreatment step for 2*U disposal
in HLW glass (Sect. 4.1). Product characteristics are shown in Table 4.8.

The high-quality glass fusion melt process can accomplish the same tasks and produce a waste form
acceptable for a WIPP- or YM-type repository. High-quality waste glasses (typically with high silica
contents) require more complex processes and have significantly lower concentrations of uranium in the
product. This process would not be suitable for dilution of **U to 12 wt % ***U in ***U (nonweapons ***U)
that is stored for potential future uses because of the difficulty in recovery of uranium from such a glass.

Product characteristics are shown in Table 4.8.
56

Table 4.6. Summary: Convert >*U to borate fusion melt

 

Application
Acceptable **U feeds

Method to convert to non-weapons-usable
233
U

Disposal site
Criticality control strategy

Technical description

Assessment

Advantages

Disadvantages

Evaluation

Storage, disposal
All

Isotopic dilution

WIPP-type repository
Isotopic dilution

Excess ***U and DU oxides are dissolved in borates [boron
oxide (B,0O;) or borax (Na, B,0-)] in an induction furnace
to create a fusion melt. The melt is cooled to a solid and
packaged.

Potentially the simplest, lowest-cost option to isotopically
dilute ***U oxides with DU oxides.

Can dissolve uranium oxides from sliced CEUSP storage
containers.

Product form would allow future recovery of the
isotopically diluted **U or potentially valuable decay
products (**Th for medical applications).

The process may also be used as a feed preparation step to
prepare *°U for disposal in HLW tanks (see Sect. 4.1).

Low-quality waste form (boron oxide soluble in water)
suitable for disposal in a WIPP-type facility, but not in a
YM-type facility.

Some additional process development work required.

This is a viable option and may be the low cost option if
1sotopic dilution is the primary requirement. There is the
potential to modify the process to produce a better waste
form.

 
57

ORNLDWG98C-373

 

 

 

 

 

 

 

 

 

 

 

 

Fusion Melt DUO,
233) 0, B,O; or Na,B,0,
# Vertical Tube Furnace
Crucible ®* Option = SlitStorage
Can and Put 233U
OxideandStorage
Induction Can into Drum
Furnace # Borate Chosen For:
-FastDissolution
\" -HighUranium
\}\\% Loading
Weld Lid
< UraniumBorate Solid
Storage

       
  
     

SR
SIS

£

   
 
 

   
 

e

  
    

SRR E: N %fih&
Existing 23U Waste WIPP-type
StorageArea GeologicalRepository

Fig.4.6.Borate-fusion melt for >3*U long-termstorageordisposal.
58

Table 4.7. Summary: Convert 2°U to glass by fusion melt

 

Application
Acceptable **U feeds

Method to convert to non-weapons-usable
233
U

Disposal site
Criticality control strategy

Technical description

Assessment

Advantages

Disadvantages

Evaluation

Storage, disposal

Clean and CEUSP materials. LWBR **°U as a feed may
be unacceptable because of its thorium content.

Isotopic dilution

WIPP- and YM-type repository
Isotopic dilution

Excess “**U and DU oxides are dissolved into a molten
glass frit (metal oxides) in an induction furnace to create a
fusion melt. The melt is cooled to a solid and packaged.

Final waste form is acceptable for any disposal site.

High confidence that process will work based on HLW
glass making experience.

Waste form development is required to develop high-
uranium waste glass that meets YM WACs. Uranium
loadings in the glass strongly impact process and disposal
cost.

Process development work is required.

It 1s unclear whether a practical thorium glass with high
waste loadings can be developed.

This is a viable option because of the large experience base
that exists with glass waste forms. There are major cost
uncertainties. These uncertainties derive from two
technical uncertainties: (1) maximum allowable uranium
and thorium loading in the glass and (2) chemical reaction
rates between glass frit and the uranium.

 
59

ORNLDWG98C-370R
Glass Melting

 

DUOj
233U0, * * Glass Frit

 

 

 

   
   

——— Crucible

 

 

 

 

 

Weld Lid

Disposal

 

YM-type G eological
Repository

 

 

Fig.4.7. Glass-fusionmelt(alkaliglass) optionfor dispositionof 2.
60

Table 4.8. Product volume and mass per kilogram of >°U

 

 

 

 

Volume Mass
(L/kg *°U) (kg/kg *°U)
Processing option <12 wt % **U  <0.67° wt % **U <12wt % U <0.67° wt % ***U
Boron oxide” 1.98 44.6 12.5 282
Quality glass 5.0 115.5 19.8 446

 

“The required isotopic dilution for criticality control is 0.66 wt % **°U in pure ***U. If DU with
0.2 wt % *¥U is used, the final concentration of **U is 0.53 wt %. Some of the >**U must be used for
criticality control of the ***U in the DU.

’The B,O, fusion melt is assumed to be 80 wt % UO, and 20 wt % B,O;. The respective densities of
B,0;, UO;, and a fusion melt with 80 wt % UO; (estimated) are 2.46 g/cm’, 7.29 g/cm’, and 6.32 g/cm’.

4.4.1 Process Descriptions

In all fusion melt processes (ORNL July 27, 1995), the ***U oxide is mixed with DU oxide powder and
solvent metal oxide powder and melted in an induction furnace to produce a material in which the **U
cannot be separated from ***U by chemical or mechanical means. The solvent metal oxide dissolves the
uranium oxide into a molten solution. The use of solvent oxides reduces fusion melt temperatures from
2,500°C for pure UQO, to temperatures between 800 and 1,200 °C. In a borate fusion melt process, the
solvent metal oxide is B,O; or Na, B,O,. Other oxides may be added to lower glass viscosities and, thus,
operating temperatures. The loading of uranium oxides may exceed 80 wt %. In the glass fusion melt
process, a glass frit containing alkali oxides and silica is used to produce a high-uranium-content glass with

uranium oxide loadings up to 50 wt %.
4.4.1.1 Glass Fusion Melt Process Description

The glass fusion melt process is shown in Fig. 4.7. The ***U material to be processed is removed from
the storage wells and transported in a shielded carrier to the processing hot cell. The material is then
processed in limited-sized batches for criticality control. Inside the hot cell, the material is inventoried for
accountability control. After inventory measurements, the containers are opened, and the contents are

removed. Many of the containers, such as the CEUSP monoliths, require equipment to cut open the
61

container and to drill into it or otherwise to extract the contents. After any large chunks are removed from
the containers, the chunks are crushed into granules before the material is sent to a ball mill to be
powdered. Any of the oxide forms in which the uranium is stored (UO,, U;Oq, etc.) can be used. Metallic
uranium and low-oxide uranium may require conversion to oxide ( U;Oy ) before blending. If so, this is
affected by heating the uranium in a furnace under an air or steam atmosphere. The nominal processing
rate is 1.2 kg of fissile uranium per day, assuming a three-shift-per-day operation and a 4-h cycle per batch
using one melt-blend furnace. However, the processing rate could be increased by installing parallel
systems.

After the U material is converted to an acceptable powder form, it is blended with DU and a
specified solvent oxide mix consisting largely of alkali-metal oxides and silica. Since the melt-product
crucible also serves as the product container, it should be sized for the final product volume. As such, the
crucible will not hold all of the material in loose, flowable, powder form. Several powder additions must
be made during the processing of a single batch of material. The initial charge does not contain the DU.
With the onset of melting, DU oxide is periodically added while the mix is stirred. The final process
temperature will be adjusted to achieve complete melting for the type *°U product glass being processed.
No significant chemical reactions occur during mixing and melting other than dissolution of the uranium
oxide. The isotopic content of the mix is verified by sampling. After the contents are melted and mixed,
the furnace is turned off, and the contents are allowed to solidify in the melt crucible.

To provide double containment for storage and shipment, the crucible containing the product is sealed
inside an inner container by welding. The inner container is then placed inside a second container, which is
also sealed by welding.

The glass fusion melt process adds alkali-metal oxides, silica, and DU to final product. The isotopic
dilution of 1 kg of **U to non-weapons-grade uranium by the dry melt process will produce ~19.8 kg of
oxide product (containing 8.407 kg of U and 9.9 kg of alkali metal oxide and silica) having a volume of
5 L. Blending 1 kg of ***U to eliminate long-term criticality will produce 446 kg of oxide (189 kg U as
U,O; and 214.4 kg of alkali metal oxide and silica) with a volume of 115.5 L. The actual waste volume
per kilogram of ***U will be higher for those feeds containing impurities (thorium oxide, cadmium, and
gadolinium oxides). Some impurities, such as thorium, may significantly increase waste volumes because

of solubility limits of these impurities in glass.
4.4.1.2 Boron-Oxide Fusion Melt

The boron-oxide fusion melt process is similar—except that some steps may be simplified, melting
times may be reduced, and waste loadings may be increased. This follows from choosing a metal oxide

solvent based on its properties to dissolve uranium oxides quickly at relatively low temperatures (which
62

minimize off-gas complications) rather than choosing a metal oxide solvent based on glass performance
requirements. The option may minimize or eliminate the need to grind the ***U feed materials. Second, the
process may be applicable to LWBR *°U. Thorium oxide is not readily soluble in traditional glasses. It is
an option herein. In this context, the ThO, concentration will be low in the final product because of the
addition of large quantities of DU.

As a part of several programs in Germany and the United States to develop chemical core catchers for
nuclear power reactors, significant work (Dalle Donne et al. 1978; Forsberg et al. 1997) has been done on
the dissolution of uranium oxides in different borates. Core-melt accidents are among the most serious
nuclear reactor accidents. In a serious accident, the core itself melts, then melts a hole through the pressure
vessel, and next melts a hole through the building containment floor that allows release of radioactivity to
the open environment. Core-melt accidents are difficult to control because the primary material in a core
melt is molten uranium dioxide at a temperature of several thousand degrees Celsius. In a chemical core
catcher, a specially selected compound is placed under the reactor core. When the core debris reaches the
floor, the decay heat melts the specific compound, and the uranium oxides are dissolved into the compound.
The compound is chosen to have a high uranium loading and melt at a low temperature. The liquid with
the dissolved uranium rapidly spreads out over the reactor building core in a geometry that allows rapid
cooling. The major compounds that have been investigated for this application include B,0O;, Na, B,O,,
and lead borate (B,052PbQO). The requirements for a chemical core catcher (high uranium solubility in
liquid, fast dissolution, and low temperatures) are essentially identical to those needed for fusion melt
isotopic blending.

The reactor core-catcher data suggest the potential for major process simplifications which would have
a large impact on the cost of processing CEUSP and LWBR material. The CEUSP ***U was solidified
inside its stainless steel container and is partly attached to the container walls. Removal of the **U from
the container would be a complex, expensive, mechanical operation. If the dissolution kinetics are
sufficiently fast, a relatively simple batch process operation would be possible. The furnace with
disposable inner liner would be loaded with the borate flux and some DU oxide. The ***U container would
be cut into several pieces and dumped into the liner, the furnace would be heated to operating temperature,
and the **U would be dissolved into the molten mixture from the container pieces. After dissolution of the
#¥U, additional DU would be added to the furnace, the dissolution process would be completed, the melt
would be cooled, and the solid product (with container parts) would be packaged as one piece. This type of
operation is inconsistent with production of high-quality glass but may be feasible if the product does not

need to meet YM-type waste-glass performance specifications.
63
4.4.2 Uncertainties

Work is required to develop an uranium glass process. Many uranium glasses have been produced and
recorded in the literature. However, these glasses have not been evaluated for their performance under
repository conditions. Furthermore, only limited engineering data exists on these glasses. Because of the
broad technical experience base with waste glasses, including HLW glasses containing several percent
uranium, a technically acceptable glass can obviously be produced. However, the economics (processing,
transport, and disposal cost) strongly depend upon obtaining a high uranium waste loading in the glass to
minimize process equipment size and final glass volumes.

Work is required to develop a boron-oxide fusion melt process. The two borates of primary interest are
B,O; and Na, B,O,. The B,0O; should produce a product with a higher loading of uranium. Borax may
have somewhat faster dissolution kinetics. Additional studies and thermodynamic and chemical kinetic data
on uranium dissolution in fusion melts is required to predict (1) processing time and (2) if grinding of the

feed is required in the process.
4.4.3 Conclusions

Both variants of the fusion melt option are workable and viable. Additional technical work is required
to reliably predict costs. Based on available experimental data, the borate fusion melt technology is

potentially the lowest cost technology to isotopically dilute ***U with DU.

4.5 DRY GRINDING, BLENDING, AND SINTERING (STORAGE AND DISPOSAL)

In the dry-powder-blending process (Table 4.9; Fig. 4.8), ***U oxide and DU oxide are ground to fine
powders, mixed together, and consolidated to produce a ceramic-type material from which the **U cannot
be separated from the **U by chemical or physical means. The final product could be sent to long-term
storage, a WIPP-type facility, or a YM-type facility. To achieve these goals requires that the process must
fuse the finely ground mixture of oxide particles and should promote the interdiffusion of DU and ***U
atoms, irreversibly intermixing them on an atomic scale to a point that only an enrichment plant could

separate them again.
64

Table 4.9. Summary: Dry grinding, blending, and sintering

 

Application
Acceptable **U feeds

Method to convert to non-weapons-usable
233
U

Disposal site
Criticality control strategy

Technical description

Assessment
Advantages

Disadvantages

Evaluation

Storage, disposal
Clean, LWBR, CEUSP

Isotopic dilution

All
Isotopic dilution

Excess ~°U oxides are ground and mixed with DU dioxide
until the average particulate size is ~0.5 um. The powder
is pressed and then sintered by heating to a high
temperature. The product is a monolithic solid. If the *°U
and DU oxides have the same chemical compositions,
several particle consolidation processes can be used. If the
*3U and DU oxides have different chemical compositions,
a high-temperature sintering process is needed that
isotopically mixes the two isotopes by solid diffusion.
Fine-powder grinding and powder mixing is required for
reasonable sintering times and temperatures.

This may be a low-cost process.

The process would be more complex to use for CEUSP
feeds because of the presence of cadmium that will
probably volatilize in the sintering furnace.

The process uses fuel-pellet fabrication technology. This
1s a moderately complex technology. However, the
product requirements for the dry-blend option are much
less than those required for fuel fabrication; thus,
simplification of the process should be possible.

The process produces very fine *°U particulates (~0.5 um)
that are a significant inhalation hazard. Special safety
procedures would be required compared to most other
disposition options.

This potential option to isotopically blend all **U with DU
1s a reasonable one.

 
65

ORNL DWG98C-369R3

Dry Blend
DUO; (Match Chemical

*U0; * h Form of 2°U)

Size-Reduce/
Blend

 

 

 

 

 

 

 

  
  

Solidification

Agent
; (Option - Same
Containerfor
Grinding,
Consolidation,
and Packaging)

No Action

 

 

 

 

 

 

 

      

Storage (Use/Throw)

 

   

.................... M
%“ E\\\\:\:&} h:,\“s\\i; - B
Existing 233U Waste WIPP or YM-type
StorageFacility Geological Repository

Fig.4.8. Simple dry-blendprocessing for Py long-termstorageordisposaloptions.
66
4.5.1 Process Description

The dry-blend processes are similar to the fabrication methods for LWR fuel pellets. Several
equipment variants are possible. Two variants are described herein. The first variant uses essentially the
same equipment as is used for fuel fabrication. The second variant modifies the equipment to simplify the
process.

The **U to be processed (ORNL July 19, 1995) is removed from the storage wells and transported in a
shielded carrier to the processing hot cell. A bagless loading procedure is used for contamination control.
The material is processed in limited-sized batches for criticality control. Inside the hot cell, the material is
inventoried for accountability control. After inventory, the containers are opened, and the contents are
removed. Many of the containers, such as the CEUSP monoliths, require equipment to cut them open and
to drill or otherwise extract the contents. After any large chunks are removed from the containers, they are
crushed into granules before the material is sent to a ball mill to be powdered. Other grinding and mixing
technologies may be used. Any of the oxide forms in which the uranium is stored (UO,, U;Oq, etc.) can be
used. Metallic uranium and nonoxide uranium are converted to oxide (U;QOg) by heating in a furnace under
an air atmosphere before blending. The nominal processing rate is 1.2 kg of fissile uranium per day,
assuming an operation of 3 shifts per day and a 4-h cycle per batch. The batch size is chosen to be 200 g
to ensure criticality control. However, the processing rate could be increased by installing parallel systems.
Material in the feed other than uranium oxides, such as cadmium, gadolinium, or thorium oxide, will
remain with the uranium oxide throughout the process and be present in the final product.

Blending begins by adding a weighed amount of ***U oxide powder with a predetermined quantity of
DU dioxide powder in a mixing vessel. DU dioxide for blending is prepared outside the hot cell. Mixing
the powder 1s accomplished by rotating the mixing vessel in a tumbling apparatus for a prescribed length of
time. An organic binder (to aid in the subsequent pellet-making process) is added and blended with the
powder during the mixing operation.

Cold-pressing of the mixed oxide powder is accomplished by an automatic press. Blended powder is
fed into a cavity and pressed by a piston. The ensuing pellets are to be limited to a mass of 57 g or less and
a maximum diameter or height of 3 cm, and the nominal fissionable material in each pellet will not exceed
1 g. The pellets are then ejected from the cavity into a sintering tray. The pressing creates a sinterable
product.

The uranium oxide pellets are placed in a sintering oven and heated to a temperature that is sufficient

such as to vaporize the organic binder and to heat the pellets to near melting. The sintering process
67

converts different uranium oxides to uranium dioxide (UQO,). The pellets reduce in size as the volatile
materials are driven off. The finished pellets will be ~90% of theoretical density (9.86 g/cm’ for UO,) of
the mixed oxide. The sintering also increases the particulate size and reduces the inhalation hazard
associated with small particulates. After the pellets are sintered, they are allowed to cool and are then
transferred to an area for nondestructive analysis to verify their fissile content.

The dry-blend process does not introduce any additional material into the product other than the DU.
Isotopic dilution of 1 kg of ?**U by the dry-blend process will produce 9.5 kg of non-weapons-usable UO,
(8.407 kg U) product having a pellet volume of ~0.96 L. Blending 1 kg of **U with DU to eliminate long-
term criticality will produce 214 kg of uranium oxide (189 kg U) product having a pellet volume of
~21.7 L. The density of the pellets can be partially controlled by the pellet’s pressing and sintering
conditions. However, the actual waste volume per kilogram ***U will be higher because of the presence of
other material (thorium, cadmium, and gadolinium oxides) in the ***U material, any additional DU required
to dilute **°U contained in the **U material, and any DU added to accommodate measurement uncertainties.

After sintering, the pellets are placed in the inner containment vessel (can), which is sealed by welding.
The can is then enclosed in an outer container to provide double containment for storage and shipping.
Off-site transportation is carried out in shielded over-packs that meet U.S. Department of Transportation
(DOT) requirements.

There are several other equipment configurations to the simple dry-blend option (R. Harmon,

C. V. Smith, and R. Henry 1998). One option being examined for dry down-blending of HEU with DU
consists of grinding and mixing *>UQO; and DUQ; in a dual drive planetary mill with alumina ball grinding
media for ~75 minutes. This option has both advantages and disadvantages. The grinding media remains
with the uranium oxides and occupies ~35 vol % of the final product. The use of a grinding media—rather
than separate grinding and mixing steps—may allow the grinding and blending container to be used as a
final product container provided that the container can be compressed and survive sintering furnace

conditions. This avoids transfer of the fine powder during operations.
4.5.2 Variant Isotopic Mixing Methods with Dry Blending

There are three dry blend options. The options, their advantages, and their disadvantages are described
herein. The option described previously was chosen as the baseline dry-blend option because of identified

limitations in the other dry blend options.
68
4.5.2.1 Dry Blending Only

Currently, it is uncertain whether powder mixing of **U and DU in the same chemical form assures
i1sotopic irreversible mixing. If the particle sizes are large, many types of commercial equipment exist to
separate powders with different densities. There is almost a 2% difference in the densities of **UQO, and
*BUO,. At small particle sizes (<1 pm), there is the potential of self-separation by alpha-initiated
Brownian motion. During alpha decay, a high-energy alpha particle is emitted. With small particulates,
the recoil from alpha decay can move the **UQ, particle. In a mixture of *UQ, and ***UQ, particulates,
the lighter 2*UQ, with its alpha-driven motion may preferentially work its way to the top of a powder
mixture. This type of potential separation mechanism does not exist when mixing **UQO, and DUO,
powders to create LEU because of the relatively low radioactivity of each uranium isotope.

For disposal in a YM-type repository, dry blending by itself does not produce an acceptable waste
form. The current waste form requirements (DOE, August 30, 1994) prohibit waste forms containing fine

particulates. Dry blended solids could be put into storage.
4.5.2.2 Dry Blending and Consolidation

To avoid self-separation mechanisms, powder mixtures can be consolidated into a matrix after
blending. The powders can be sintered into pellets at relatively low temperatures or incorporated into a
matrix such as cement. Consolidation also minimizes risks from respirable particles during storage,
transport, and disposal. Some disposal sites (e.g., YM) prohibit fine particulate waste forms.

To assure irreversible isotopic mixing, identical chemical compositions for U and DU are required to
avoid the potential for separation of the two isotopes by selective chemical dissolution of one species or the
other species with appropriate dissolution agents. Dry-blend processes that mix powders followed by

consolidation would be difficult to use with CEUSP material and impractical to use with LWBR material.

e CEUSP*’U. The CEUSP material contains gadolinium (a rare earth) and cadmium oxides.
Gadolinium oxide in UQ, is used as a burnable neutron absorber in nuclear fuels. Rare earths are
major fission products in SNF. This type of system has been studied in detail. Oxidation and
dissolution studies (Hanson July 1998; McEachern and Taylor 1998) have shown that the addition
of rare-earth oxides—such as gadolinium oxide—to UQO, significantly alters UO, oxidation and
dissolution rates under a wide variety of conditions. It is expected that this phenomenon would
also be true for other uranium oxide species. To blend CEUSP material would require that the DU
oxides have the same concentrations of cadmium and gadolinium oxides as the CEUSP material
and have the same physical and chemical forms (sintering temperatures etc.) to avoid separation by
selective dissolution of one material or the other.
69

e LWBR?”U. The requirement that the **U and DU have the same chemical form applies to the
LWBR material. However, the LWBR feed contains only 1 to 12 wt % **U in high-fired ThO,
and must be matched to a comparable feed. The DU (for blending purposes) would contain
1-12 wt % DU in high-fired ThO,. Because of the low concentrations of uranium in the thorium,
large quantities of ThO, would be required to make the DU feed. The LWBR **U contains
~350 kg of U incorporated in 14 t of ThO,. To isotopically dilute the ***U to non-weapons-
usable material would require DU in about 120 t of ThO,. To eliminate criticality concerns, over
2,000 t of ThO, with DU would be required. This implies a large dry blending facility to handle
the ThO, masses.

4.5.2.3 Dry Blending with High-Temperature Sintering

If the **U feed has variable or complex chemical compositions (CEUSP and LWBR ***U), dry
blending with high-temperature sintering does allow the use of normal DU oxides to isotopically dilute the
2¥U. With this dry-blending variant, solid diffusion during high-temperature sintering causes isotopic
blending. Fine grains of DUO, and the ***UQ, diffuse together into larger single crystals. There is
applicable industrial experience. In the fabrication of LWBR fuel pellets, ThO, and **UQ, powders were
blending, pressed into pellets, and sintered at high temperatures until the thorium and uranium interdiffuse
to create a homogeneous Th-U oxide. This does require fine-powder blending (~0.5 pm). Long sintering
times (hours), and high sintering temperatures (~1800°C) assure diffusion of the DU oxides into LWBR or
CEUSP feed materials.

There are two specific types of complications using this approach with the CEUSP **U. At the
required sintering temperatures, much (perhaps all) of the cadmium oxide may vaporize into the off-gas
system (sublimation temperature of cadmium oxide is ~ 1500°C) with unknown effects on the furnace
components. Second, for a YM—-type repository, RCRA (chemically hazardous) materials are not accepted.
The cadmium must be (1) shown to be a beneficial material for criticality control and thus not regulated as
a RCRA material, or (2) removed from the feed, or (3) shown that it does not leach into groundwater using

the EPA TCLP waste leaching test.
4.5.3 Safety Issues with Dry-Blend Options

The dry-blend options may require some additional safety precautions. The primary hazard of *°U is
the internal radiation dose upon inhalation. The risk from inhalation is primarily dependent upon the
particulate size. Only small particulates (<1 um) can easily be inhaled into the human Iung. This process
requires that the 2’U oxides be reduced in size to ~0.5 um—the size range of high-hazard particulates.

With the dry-blend process, there is also the danger of creating pyrophoric powders with certain feeds.
Many materials, such as UO,, are pyrophoric in fine powder form; thus, appropriate process safety

measures must be implemented. If the UO, is sintered, the final product is inert and is not pyrophoric. If
70

the blended powder is to be sent to long-term storage (not disposal), it is required that the powder be
converted to a nonpyrophoric form such as U;O; (DOE December 1998). The storage standards for both
*¥U and U (Cox July 1995) do not allow storage of pyrophoric powders.

4.5.4 Mechanisms of Isotopic Dilution with High-Temperature Sintering and the Implications on
Process Design

The applicable experience base for this option is fuel fabrication—particularity LWBR fuel
fabrication. However, fuel fabrication is relatively complex because of the need to make nuclear fuel
pellets which meet multiple requirements. For dry blending, simplified flowsheets may be viable because
(1) the primary criteria is isotopic dilution and (2) most fuel fabrication requirements are irrelevant.
Process simplification may be possible.

The grinding and dry-blending processes for the storage and disposition process are similar to the
fabrication methods for LWR fuel pellets, (Frankhauser et al., 1967), (Leitnaker, Smith and
Fitzpatrick, 1972). Since the final product tolerances for storage and disposition materials are not as
severe, several equipment and processing variations are possible. Two processing scenarios are compared
herein. The first scenario uses essentially the same equipment and process steps that are used for fuel
fabrication. The second scenario modifies the equipment and eliminates steps to simplify the storage and
disposition process and reduce its costs.

Both scenarios require that the **U oxide and DU oxides be ground to about an average diameter
0.5 um with a surface area of ~8.0—9.5 m*gm. In the case of fuel fabrication, the fine powder mixture is
ground and blended in five to six steps or passes. Historically, this grinding of the oxides to such a small
particle size was accomplished by using a sequence of two different sets of equipment (Belle et al., 1976).
After being milled, the final “micronizing” is done in a pulverizer. In the case of the storage and
disposition process, we would like to reduce the number of steps and the associated equipment and
handling. Modern mills may be capable of achieving these particle sizes in one step.

The pressing of these powders into “green” pellets before sintering is essential to meet the goal of
achieving high theoretical densities in the final product in order to ensure that the fuel pellets have good
thermal conductivities. To achieve this in fuel fabrication, the powders are agglomerated to aid handling,
treated with binding and mold release agents, and cold-pressed at around ~35,000 psi. This results in a
pressed “green” pellet that can be handled and that has tightly controlled dimensions. However, this
treatment also requires that there are necessary preheating treatment stages in the sintering process that

removes the added binding and release agents. These process steps require sequentially heating the pressed
71

green pellets to ~350, ~600 and over 900°C under a CO, atmosphere in order to decompose the reagents,
to react with the residual carbon, and to volatilize the remaining residues, respectively. In the case of the
storage and disposition process, almost all of these steps should be by-passed.

However, there are some aspects of powder pressing that are relevant to the storage and disposition
process. At these pressures, the contact points of the powder particles are stressed to the point that their
crystalline structures undergo force dislocations, the particles are plastically deformed, and their surface
contact areas are increased (Olander, 1976). This gives the subsequent sintering processes a head start,
accelerating the growth of the particles into coarse grains and facilitating the interdiffusion of atoms across
the particles’ increased contact boundaries. For the storage and disposition product, these coarser gains
correspond to lower leachabilities. Also, the greater initial particle-particle contact areas will accelerate
the atomic mixing between the ***U oxide and DU oxides. Therefore, a modified version of the powder-
pressing treatment may be included in the storage and disposition process.

After pressing, both the green fuel pellets and the green storage and disposition products are ready to
be sintered. The questions are, For how long, and at what temperature? Grain growth aside, the attainment
of homogeneous isotopic dilution of U oxide in DU oxides is the controlling issue. The atoms in the
*¥U oxide and DU oxide particles must have time to diffuse into one another. The self-diffusion
phenomenon is depicted in Fig. 4.9.

To achieve a 10:1 dilution of ***U in DU oxide, the mean self-diffusion distance, <r>, must exceed the
radius of the **U oxide particle by 2.61 times. Between 1450 and 1800°C, the energy of activation for
uranium self-diffusion is 88 kcal/mol (Belle 1961). Using these data, Table 4.10 calculates and
extrapolates the diffusion self-diffusion coefficients and <r> for urania particles between 0.5 and 10 um.

As the temperatures increase, so do the diffusion coefficients, and the times for atoms to reach the
2.61 - r are reduced. At the fuel fabrication conditions of 1,800°C and 0.5 um, it takes about ~ 14 h for
homogeneous isotopic mixing. With 5 um particles at 1,800°C, it would take 1,436 h. If we raise the
temperature to 2,200°C, the 5-um particles would take ~45 h to achieve isotopic homogeneity at
10:1 dilution. This would probably require an induction furnace under a reducing atmosphere of argon
with 5% hydrogen. Note also at temperatures above 1,750°C, the volatility of UO, may become a control

1ssue.
72

ORNLDWG99C-74

Self-Diffusion Time to Achieve
10:1 Atomic Dilution

 

o O
o\ £
=\ v, I3
T % /O

® 3‘;,/’\A
f
O/' < "/\:\(b
Tll‘mas't‘('\6
<r>= \D(T)-t,
(where:)

<F > = average random displacement,

D(T) = self-diffusion coefficienta s afunction of
T =temperature,

t = time.

Fig.4.9.Self-diffusion ofuraniuminuraniumoxide.
73

Table 4.10. The time required for the tenfold dilution of atoms, when the mean self-diffusion
distances, <r>, of the uranium atoms UQ, are 2.16 times the original particles' radii

 

Time (hours)

 

Original particle size (um)

 

Uranium oxide
self-diffusion

 

Temperature coefficient,
°0 Radii 0.5 0.8 1 2 5 10 (cm?/s)
1,400 2,375 6,080 9,500 38,000 237,500 950,001 1.364E-15
1,450 1,102 2,820 4,406 17,625 110,159 440,636 2.941E-15
1,500 534 1,366 2,134 8,537 53,357 213,428 6.072E-15
1,550 269 688 1,076 4,303 26,893 107,570 1.205E-14
1,600 141 360 562 2,249 14,059 56,236 2.305E-14
1,650 76 195 304 1,216 7,602 30,408 4.262E-14
1,700 42 109 170 679 4,241 16,963 7.640E-14
1,750 24 62 97 390 2,435 9,739 1.331E-13
1,800 14 37 57 230 1,436 5,744 2.256E-13
1,850 8.7 222 34.7 138.9 868 3,472 3.732E-13
1,900 5.4 13.8 21.5 85.9 537 2,149 6.032E-13
1,950 3.4 8.7 13.6 543 340 1,358 9.540E-13
2,000 2.2 5.6 8.8 35.1 219 876 1.479E-12
2,050 1.4 3.7 5.8 23.0 144 576 2.249E-12
2,100 1.0 2.5 3.9 15.4 96 386 3.362E-12
2,150 0.7 1.7 2.6 10.5 66 262 4.941E-12
2,200 0.5 1.2 1.8 7.2 45 181 7.150E-12

 

Note: The diffusion coefficients and times for temperature above 1800°C are extrapolated beyond the
range of data used to establish the activation energy of uranium self-diffusion in UQO,.

Considering only self-diffusion may be conservative because there are other transport mechanisms at
work at the higher sintering temperatures. In the ~14-h sintering period used in fuel fabrication, the
0.5-um particles merge and grow into grains >10 um. There is a driving potential to reduce the high-

energy surface areas of the fine particles that forces them to merge and grow into larger particles with a
74

much lower surface to volume ratios. Without external pressure, the mechanisms that facilitate this grain
growth include (1) the more rapid diffusion of atoms on the surfaces of the particles and (2) the
vaporization of UQ,, followed by its condensation at the very narrow particle contact points. At
temperatures above 1750°C, the kinetics of these mechanisms could be fast enough to preclude the
necessity for cold-pressing the powder mixtures. Therefore, it may be possible to achieve the required
1sotopic homogeneity at reasonable temperatures, times, and particle sizes without having to press the
mixtures before sintering.

The experience in fabrication of the LWBR fuel pellets is particularly relevant. The **UO, and ThO,
powders were ground to ~0.5 um. The pellets were pressed and sintered as ~1790°C for 12-14 h. The
photomicrographs show large homogeneous UQO,-ThO, grains. The homogeneous characteristics of these
pellets suggest that the diffusion calculations are conservative and that one or more mechanisms described
previously accelerate mixing on an atomic scale.

The previous considerations indicate that processes for dry blending of ***U with DU for storage or
disposal may be considerably simpler than those required for fuel fabrication with resultant implications for
processing costs. As shown in Table 4.10, various combinations of particle sizes, temperatures, and
sintering times can achieve isotopic dilution of Z*U. Because nuclear-grade fuel pellets are not required, it
may be feasible to press at relatively low pressures uranium oxides into a disposable sintering-boat and
avoid multiple handling operations. Other possible simplifications include the development of a “liquid”
sintering reagent to be added to the **U oxide and DU oxide mixture to accelerate their mixing at lower

temperatures. Ideally, this fluxing reagent would disappear, dissolving into the urania oxides.
4.5.5 Conclusions

The process is potentially viable for isotopic dilution of clean, LWBR, and potentially CEUSP ***U and
would produce a product acceptable for storage or disposal in any repository. Further process development
may be required to (1) simplify the process when nuclear grade fuel pellets are not required and (2) address

specific issues associated with cadmium in the CEUSP ***U feed material.

4.6 CHEMICAL DILUTION: WASTE THRESHOLD (DISPOSAL)

If fissile materials are sufficiently dilute, they may be disposed of as other radioactive wastes. The
option exists to dilute ?*U in other materials to low concentrations to eliminate it as a weapons usable

material and then dispose of it as waste. This is summarized in Table 4.11 and Fig. 4.10.
75

Table 4.11. Summary: Chemical dilution: waste threshold

 

Application
Acceptable **U feeds

Method to convert to non-weapons-usable
233
U

Disposal site
Criticality control strategy

Technical description

Assessment
Advantages

Disadvantages

Evaluation

Disposal
All

Chemical dilution of *’U with fissile mass limits per
container.

WIPP-type repository
Mass limits (optional isotopic dilution)

There are two variants.

Coprocess:  The **U is mixed with TRUW for disposal
at a WIPP-type facility with the **U being
chemically diluted to <0.15 wt % **U.

Stand-alone: The LWBR **U is packaged in special
containers for direct disposal in WIPP (pipe-
and-go variant). The **U mass is limited to
<200 g/55-gal drum.

Low cost.

There are policy questions for the stand-alone variant:

(1) does this option meet the Spent Fuel Standard and

(2) does the United States want to advocate nonisotopic
methods for conversion of weapons-usable ***U to non-
weapons-usable “*U. The Spent Fuel Standard can be met
with the coprocessing option if there is sufficient dilution
of the #*U with other wastes.

There are institutional uncertainties associated with mixing
CH TRUW and **U.

This is likely to be the lowest-cost disposition option for
LWBR material—particularly if the operations can be
integrated with ongoing CH TRUW processing operations
at Idaho. It is potentially the lowest cost option for other
233U.

For the pipe-and-go variant, there is a policy question of
whether this option fully meets the Spent Fuel Standard.

 
Chemical Dilution

 

233U02

  

Storage

 

e

e
R

ST
s A

il LT g

T

St

N
ot \}\\E

9

7

%, 4
2, !

KX
-

B
%7
w0

o
b

A
{

 

Existing 233U Waste
Storage Facility

 

 

ORNL DWG98C-377R3

DUO, (Variant)
Waste (Optional)

SolidificationAgent
* | (Optional)

s Chemical Mixing
* No Isotopic Mixing of

Uranium Isotopes

* No Radiation Barrier

  
 

 

 

Disposal

 

 

WIPP-type
Geological Facility

Fig.4.10. Chemicalwaste thresholddispositionoption for 233UinaWIPP-typefacility.
77
4.6.1 Definition of Waste Threshold

Radioactive wastes contaminated with low concentrations of fissile materials can be disposed of as any
other similar radioactive waste. An example is HLW—which contains plutonium—that will be disposed of
in YM. High-assay fissile materials can be an economic waste (no value or future use), but fissile
materials can not be disposed of as traditional radioactive wastes because of the potential for (1) nuclear
criticality and (2) use of such materials in nuclear weapons. Because fissile materials can be used for
weapons, fissile materials have safeguards requirements that do not apply to typical radioactive wastes.
The requirements for storage, transport, and disposal of traditional radioactive wastes—such as HLW—are
significantly different than those for fissile materials. The definition of what is waste and what is fissile
materials 1s defined as the waste threshold. This definition defines what material is to be disposed of as
fissile materials. If the concentration of fissile materials can be reduced to a sufficiently low level by
dilution with other materials to below the waste threshold, they can be treated as a normal radioactive
waste.

Historically, a clear economic division has existed between wastes and some concentrated fissile
materials. This division (the economic discard limit) was based on economic criteria for materials
containing HEU or plutonium. These fissile materials are required to manufacture nuclear weapons and
were in short supply during the cold war. Fissile material in any scrap and residue stream was considered
waste only if the cost of the recovery of the fissile material exceeded the cost of producing new material. If
clean fissile materials could be recovered from the scrap and residue stream for less than the cost of
producing new material, the stream was considered concentrated fissile material. The high cold-war
economic value of fissile materials resulted in low concentrations of fissile material in the wastes which, in
turn, minimized concerns about nuclear criticality, safeguards, and arms control. With the end of the cold
war, the United States had excess fissile materials, and the economic criteria dividing wastes from fissile
materials are no longer applicable.

DOE has undertaken studies to define the new waste threshold. The **U waste threshold study

(Forsberg, Storch, and Lewis July 7, 1998) concluded that there were three categories of materials:

»  Wastes contain sufficiently small masses or low concentrations of fissile materials such that they
can be managed as typical radioactive waste.

» Concentrated fissile materials contain sufficient fissile materials such as to warrant special
handling to address nuclear criticality, safeguards, and arms control concerns. Certain
concentrated fissile materials may have no use and are therefore excess material; however, they are
classified and managed by their material characteristics rather than programmatic demand. Thus,
even if a concentrated fissile material is not needed, it must not be managed as a waste until it
meets waste criteria.
78

» Exception case material is defined as material that should be examined on a case-by-case basis to
determine if it is waste. Many of the post-cold-war arms control and safeguards issues have not
been fully resolved for plutonium and ***U. Depending upon what is decided with the management
of those materials, decisions can then be made about the specific Z*U-containing materials. It is
likely that in many cases exception case material will be classified as waste requiring some
additional special handling—between that required for wastes and that required for fissile materials.

Three issues impact the definition of wastes, exception case material, and fissile materials.

»  Domestic safeguards. Concentrated fissile materials must be safeguarded to prevent their theft by
domestic or international terrorists. Existing and proposed waste-management processing, storage,
transport, and disposal facilities are not designed as high-security facilities to prevent theft of
weapons-usable material. To be a waste, the fissile content must be sufficiently low (enrichment or
concentration) such that it is not an attractive target for theft.

»  Arms control and international safeguards. The United States has implemented a post-cold-war
policy to encourage worldwide reductions in inventories of weapons-usable materials. The
objectives include mutually verifiable reductions in weapons-usable fissile materials with Russia
and, ultimately, other countries. As part of this policy, the United States has proposed that excess
weapons-usable material in the United States and Russia be converted into a form that makes its
recovery as difficult or more difficult than recovery of plutonium from LWR SNF. Logically, the
criterion, known as the Spent Fuel Standard, should also be applied to all waste-containing
materials; otherwise, any country could adopt waste management policies that declare concentrated
fissile materials as waste and bypass the intent of the proposed arms-control treaties to limit
weapons-usable fissile material inventories. Policy decisions are being made on the applicability of
the Spent Fuel Standard to different types of waste. Not all issues have been resolved. Most
fissile-containing wastes generated in the cold war meet the Spent Fuel Standard because they were
considered wastes only if fissile material recovery was more difficult than producing new fissile
materials. New fissile plutonium and **U were produced by recovering them from SNF or
irradiated targets.

*  Nuclear criticality in disposal facilities. Regulatory and performance requirements for disposal
sites require that nuclear criticality be avoided (1) during the operational phase of the facility and
(2) after disposal-site closure. Fissile materials may migrate in a disposal facility over geologic
time frames. Waste form composition and special facility design features are required to avoid
nuclear criticality. These requirements place limits on the concentrations and quantities of fissile
materials in waste.

Based on analysis (Forsberg, Storch, and Lewis July 7, 1998) of the previous factors, the following
definitions were proposed to define ***U-containing wastes, >’U exception case material, and ***U-

concentrated fissile material (Fig. 4.11). DOE has made no decisions on whether to adopt these

recommended definitions.
79

ORNLDWGS8C-220R2

Concentrated Fissile Material

Break in t t Safeguards
Inventory Issue Resolution Limits

 

12 wit% 233U J

Exception Case Material

(Future Management Dependent Upon Pu
Precedent and Unique Conditions)

 

Waste Defined By Three Requirements

Economic Criticality Safeguards
No Value <200 g 2%u/55-gal drum Homogeneous
or Mixture with

<1 kg »%U/m3

(Equivalent to
<200 g 233 U/55-gal drum)

or

Waste <12 wi% 23U in 238y

or

MoreRestrictive
Rules For233U
Concentrated

Wastes

<0.66 wt% 233y in 238y

Fig.4.11.Definition of 2**U-containingmaterials.
80

Waste. Uranium-233-containing material is waste if it meets three requirements: (1) there is no
existing, planned, proposed, or potential use; (2) the U (a) has a concentration of <200 g
*¥U/55-gal drum or (b) the enrichment level is <0.66 wt % ***U in ***U; and (3) the **°U (a) has an
approximately homogeneous concentration of <1 kg **U/m’ (equivalent to <200 g/55-gal drum), or
(b) the enrichment level is <12 wt % **U in Z®*U. This definition includes the three requirements
(economics, criticality control, and safeguards and arms control) that must be met to allow the
material to be handled as other radioactive wastes.

Criticality control is assured (a) by limiting the #*U concentration in the wastes (<200 g
*3U/55-gal drum) to the same concentration limit as used by WIPP and other waste management
facilities for operational safety (DOE April 1996) or (b) by isotopic dilution of ***U with ***U to a
U isotopic concentration limit (<0.66 wt % ***U in ***U) that is equivalent to 1 wt % *°U in Z*U
(Elam et al. 1997).

Avoidance of special safeguards and arms-control constraints is obtained by (a) limiting the ***U
concentration or (b) isotopically diluting the ***U to <12 wt % with ?**U to make it non-weapons-
usable material (Forsberg et al. February 1998). The requirement for an approximately
homogeneous concentration of >°U in the wastes explicitly recognizes that a small can of more
concentrated *°U in a large WP has fundamentally different safeguards and arms control
implications than if the **U is distributed through the waste. If the **U is concentrated in the
waste in a small fraction of the waste, it is easier to recover and this changes the S&S
requirements. Concentrated material is defined herein as containing >0.15 wt % **U in non-**U
materials. This concentration limit is identical to the maximum allowable plutonium content in
wastes during the cold war (waste threshold). For containers containing concentrated material, the
U content of the container is limited to 250 g/m’ (50 g/drum) with certain other restrictions.

Exception case. Exception-case material is defined as material that should be examined on a case-
by-case basis to determine if it is waste. Many of the post-cold-war arms control and safeguards
issues have only been partly resolved for plutonium and ***U. Depending upon what is decided
with the management of those materials, decisions can then be made about the specific *°U-
containing materials. For the purposes herein, exception-case material includes materials not in
the definition of wastes as described above and containing up to 12 wt % **U (chemical dilution in
non--*U containing materials).

The upper limit of exception-case category is close to the maximum concentration of fissile
materials proposed for any waste for which safeguards and arms-control restrictions may be
reduced. It is also a natural break point in the **’U inventory. There is little ***U-containing
materials between the highest assay **°U fresh fuel (12 wt % **UQ, in ThO,) and the inventories of
nearly pure **°U fissile oxides or mixtures of **U and **U.

Concentrated fissile material. All other *U materials (excluding SNF) are defined as
concentrated fissile material. These materials contain >12 wt % fissile material.
81
4.6.2 Disposal Options

The existence of a waste threshold allows disposal of U as a waste if the concentration of 27U is

sufficiently low. There are two practical waste-threshold disposition options.
4.6.2.1 Convert to Waste: Process with CH TRUW

The ***U can be chemically diluted with other materials until the product meets the classical definitions
of waste. The diluent can be clean materials or other wastes with low fissile content. This option is
applicable to all *U that is to be disposed of and is described in Table 4.11 and Fig. 4.12. The practical
option is to mix the **U with TRUW (Fig. 4.12) and then disposed of in a WIPP type facility. This option

is potentially attractive because of three factors:

» Treatment process. Construction is beginning (DOE January 1999) on the Advanced Mixed
Waste Treatment Facility (AMWTF) at INEEL to convert existing CH TRUW into a form that
meets WIPP WAC. The planned treatment process consists of several steps: (1) wastes are
unloaded from the existing containers and sorted into categories; (2) combustibles are incinerated;
(3) large components are cut into pieces and placed in drums or other containers; (4) incinerator
ash, liquids, and small particulates are mixed with cement; and (5) the cement is placed into drums
or other containers. Some of the drums and other containers will contain large components. In
these cases, the cement is poured into the void spaces to make efficient use of drum volume. The
solidified final product meets all WIPP WAC:s.

This flowsheet can be modified for U disposition. Uranium ***U is ground to a coarse powder
and added with DU oxides to the cement containing the incinerator ash, liquids, and small
particulates. The mixture is then poured into the waste drums. The **U grinding process would
be similar to that described in Sect. 4.5 on dry blending except the particle size would be larger.
This avoids the specific health and safety concerns associated with fine grinding of *U. Several
other methods to create a feed are potentially possible.

» Inventory location. The LWBR ***U and a large fraction of DOE TRUW are stored in Idaho. The
AMWTF is being built in Idaho.

e  TRUW inventory. The inventory of TRUW allows dilution of the ***U to low concentrations
without production of significant quantities of added waste. The projected final CH-TRUW
processed-waste volume at INEEL is 36,530 m’, containing 720 kg of fissile material. This
implies that the average CH-TRUW drum will contain <4 g of fissile materials/55-gal drum. In
theory, about 35 t of fissile materials could be added to the existing CH-TRUW inventory at
INEEL while meeting the WIPP WAC limit of <200 g fissile material/55-gal drum. The allowed
35 t of fissile material that can be added to this waste stream is more than an order of magnitude
larger than the total quantity of U in the inventory.
82

ORNLDWG98C-392R2

Water Cement CH TRUW

 
       
   

 

UraniumOxide

Powder Blend
(Minimal Fines)

 

 

Limited Grind/Mix

 

Large TRU Grout Mixing
Components

Grout

 

Load Debris Add Grout

  

)
o "::;:%s\

233 U Waste WI PP'type
Storage Area Facility

Fig.4.12.Chemical waste threshold 23y dispositionoption:CoprocessingwithCHTRUW.
83

Nationwide, the current and future quantities of TRUW available for mixing with **U are much
larger. In theory, ~300 additional tons of fissile material could be added to the existing and future
inventory of TRUW (317,600 m?) ultimately requiring disposal in a WIPP-type facility without
violating WIPP-type criticality limits of 200 g/55-gal drum. For most of this waste, no decisions
have been made on where it will be processed. A significant fraction is likely to be processed in the
INEEL facility. The legal agreements with the State of Idaho allow processing of TRUW from
other sites provided the processed TRUW is removed from the state within a fixed time after
entering the state.

This variant is likely to be the lowest-cost option for the disposal of LWBR ***U and a low-cost option
for other 2°U. It uses facilities that are under construction. Preliminary evaluations indicate that this
option does not significantly change those facilities and is unlikely to significantly impact the operating cost
of those facilities. It does not significantly increase the number of drums going to WIPP-type facilities.
The CH TRUW uses most of the physical volume of a waste drum. The *°U uses excess fissile “capacity”
of the waste drum, but very little of the volume of the drum because of the high densities of uranium
oxides.

The option should allow disposition of **U with higher radiation limits. Processing of this material
may require some modifications in process operations. To minimize radiation levels, two actions can be
taken.

Dilution of ®’U. The *U with higher radiation levels may require additional dilution with TRUW
to control radiation levels (see Appendix A).

e Special placement of cement containing *>U. The option exists to preferentially place the TRUW
cement containing **U with higher radiation levels in the center of waste drums and place TRUW
cement with lower radiation levels near the outside of the drum. In effect, low gamma-emitting
wastes function as shielding for the higher gamma-emitting wastes.

It may be desirable to add DU with the **U. There are two benefits. It provides a higher assurance of
criticality control and it makes recovery of the ***U more difficult. Methods for recovery of uranium in low
concentrations from cement would partially or fully isotopically mix the ***U with the DU. These benefits
are associated with clean and CEUSP **U but may not exist for LWBR **U. The unique chemical
characteristics of the LWBR material would allow its separation from cement and DU. The barrier for
LWBR ***U recovery must be the large-scale dilution with TRUW.

There are institutional uncertainties. Can two materials be co-processed and sent to WIPP under the
current regulatory and legal structures? The current legal and regulatory structure was not designed for

this option and thus there are multiple ambiguities that must be addressed.
84
4.6.2.2 Convert To Exception Case Material: Stand-Alone Disposal

The LWBR U may be disposed of as exception case material. DOE has recently developed a
strategy for disposition of plutonium scrap and residue (Voorheis April 2, 1998; DOE August 28, 1998).
This material is the plutonium equivalent of **U exception case material; thus, a similar strategy is
potentially viable for Z*U. It has been decided (DOE February 11, 1999) to send most plutonium salts and
other materials with plutonium concentrations of <10 wt % plutonium to WIPP for disposal. The amount
of plutonium per 55-gal drum is to be limited to 200 g (criticality limit). Certain other restrictions have
been proposed on the quantities of plutonium in shipment and storage to address S&S concerns. In effect,
the disposition decision for these materials requires that an intermediate level of requirements (between
those for wastes and those for pure fissile materials) be met for disposal of plutonium scrap and residue
that have plutonium concentrations in a range between traditional radioactive wastes and concentrated
fissile materials.

A parallel approach is potentially applicable to the LWBR **U inventory—with the final product sent
to a 2U waste storage area or a WIPP-type repository. This exception-case material has a **U
concentration between fissile materials and **U wastes. It is not an option for other ***U in the inventory.
The clean and CEUSP **’U would have to be diluted by a factor of 10 to meet the definition of exception-
case material. However, DU is the logical dilution agent. If DU is used, the ***U becomes non-weapons-
usable material by isotopic dilution and the waste threshold option is moot. The option is described in
Table 4.11 and shown 1n Fig. 4.11.

To implement this option, LWBR ***U would be packaged in pipe containers with <200 g of **U per
pipe container. One such container would be allowed per 55-gal drum. The waste may be managed the
same way as CH TRUW. Each package contains <200 g of essentially pure **UQO, mixed with ThO.,.
Because of the small quantities of ***U and the chemical dilution, the radiation levels from each package
will be relatively low and the material may be handled in the same way as CH TRUW. This would be
parallel to the proposed approach to manage many plutonium residues at Rocky Flats.

The LWBR material contains from 1 to 12 wt % **UQ, in ThO, vs <10 wt % plutonium in the pipe-
and-go option at Rocky Flats. On a mass basis, 12 wt % *’UQ, in ThO, is equal to 10.6 wt % ***U heavy
metal. This small fraction of the LWBR inventory is over the 10 wt % limit used to define what plutonium
fissile materials may be disposed of via this option. However, the basis used to derive the 10 wt %
plutonium limit would allow a higher concentration of **°U in the form it is found in the LWBR inventory.

The concentration limit for fissile materials using this option is imposed to minimize the risks of
85

unauthorized recovery of the fissile materials from wastes. The concentration limit is one of several
mechanisms to accomplish this task. The other primary limit is that no drum may contain >200 g of fissile
material. The 10 wt % plutonium limit was approved for plutonium in salts (McCallum May 15, 1998;
Huizenga August 17, 1998) based on estimates of the difficulty of recovery of plutonium from this matrix.
The ***U in the LWBR material (with a maximum concentration of 10.6 wt % **UQ, in ThO,) is much
more difficult to recover than 10 wt % plutonium from salts. If 10 wt % plutonium in salt is acceptable,
10.6 wt % **UQ, in ThO, should be able to meet the same requirements.

This variant allows stand-alone disposition of the LWBR ***U independent of the AMWTF. The
expectation is that this is potentially the low-cost option for LWBR ***U disposition but not as inexpensive
as mixing with CH TRUW in the AMWTF. A preliminary estimated cost to transport and dispose of a
drum of TRUW is ~$8000/drum (Appendix D). With 350 kg of ***U, the costs would be the packaging

costs plus ~$14 million for transport and disposal.
4.6.3 Special Issues with Waste Threshold Options
There are several issues associated with the waste threshold option.

» There are disagreements within the technical and policy communities on the acceptable
concentrations of fissile materials in wastes (Bunn November 23, 1998; DOE November 1997).
These disagreements are based on two considerations: (1) what is the acceptable risk of theft of
wastes containing higher concentrations of fissile materials, and (2) what are the international
implications to nonproliferation policies if the U.S. adopts a policy of allowing significant
quantities of fissile materials in the waste?

* Does the United States want to establish a precedent of converting a weapons-usable uranium
isotope to a nonweapons-usable material by any other method than isotopic dilution?
These issues directly impact any pipe-and-go option. They may or may not impact ***U disposition by

mixing with CH TRUW. With sufficient chemical dilution, the Spent Fuel Standard can clearly be met.

4.7 CAN-IN-CANISTER (DISPOSAL)

The can-in-canister option is being developed to dispose of excess plutonium. It is applicable for the
disposition of excess #*U. This option consists of (1) converting plutonium oxides (or ***U oxides) into a
ceramic form that is packaged into small cans, (2) placing the cans inside empty HLW canisters,

(3) pouring HLW glass into the HLW canisters, and (4) embedding the cans in the HLW glass. The

ceramic composition includes large quantities of neutron poisons to assure criticality control. The ceramic
86

is a highly stable waste form capable of withstanding the high temperatures within the HLW glass
vitrification process. Some of the plutonium contains HEU. It is currently planned that sufficient DU be
added to the ceramic so that any HEU fed into the process would be converted to LEU. The same strategy
would be used for **U with the addition of sufficient DU to convert it to nonweapons-usable >*U. The
option i1s summarized in Table 4.12, the option is shown in Fig. 4.13 and a drawing of the canister with
minicans is shown in Fig. 4.14.

The option is technically viable. Excess 2*U and plutonium could be processed and disposed of
simultaneously or separately in the same facilities. The plutonium immobilization program has chosen a
ceramic that contains significant amounts of uranium. The uranium in the ceramic can be U, DU, or
some combination. There are, however, constraints.

The proposed facilities for disposition of excess plutonium are being designed for CH glove-box-type
operations—not remote operations. The plutonium feed acceptance criteria for the plutonium
immobilization option allows <0.5 wt % **U in any feed material to the facility (DOE December 1998¢) or
<1 ppm **U in the **U feed material. This criteria is based on allowable radiation exposures to workers in
the facility. As a direct consequence, only the LWBR #**U could be processed by this option. The other
U materials could only be processed if the planned plutonium immobilization facility was converted from
a glovebox operation to a hot cell operation. This would multiply the total facility costs by a factor of 2 or
more and require additional years of equipment development. Consequently, it is not practical to process
other #°U feeds in the proposed plutonium immobilization facility.

If the LWBR ***U was processed in the proposed plutonium immobilization facility, there are two
options: (1) coprocess with the plutonium or (2) process in a separate campaign after plutonium
processing.

e Coprocessing. The LWBR U can be coprocessed with the plutonium. The baseline plutonium

ceramic (Ebbinghaus February 1999) has a composition of about 10 wt % CaO, 10.7 wt % HfO,,
23.7 wt % UO,, 11.9 wt % PuO,, 8.0 wt % Gd,0,, and 35.9 wt % TiO,. The LWBR **U
contains ~350 kg of **U in the form of UO, and 14 tons of ThO,. In the proposed ceramic for
plutonium immobilization, the ThO, can be used to replace some of the depleted UO, that is added
to immobilize the plutonium. This has several advantages. First, <12 additional HLW canisters

with immobilized ceramic cans would be required if thorium can replace DU on a 1 for 1 basis.
Second, no separate waste form requiring separate testing and licensing for disposal is required.
87

Table 4.12. Summary: Can-in-canister

 

Application
Acceptable **U feeds

Method to convert to non-weapons-usable
233
U

Disposal site
Criticality control strategy

Technical description

Assessment

Advantages

Disadvantages

Evaluation

Disposal
LWBR

Isotopic dilution, radiation barrier

YM-type repository
Isotopic dilution, chemical neutron poisons

Excess ~°U, DU, and other neutron poisons are converted
into a high-performance ceramic that is packaged in small
cans. The cans are placed inside empty HLW canisters
and molten HLW glass is poured over the
cans—incorporating the cans in HLW glass. This is the
same technology that is proposed for immobilization of
plutonium. The option exists to coprocess the plutonium
and U

Technically viable option using the technology being
developed for immobilization of excess plutonium. If the
33U is coprocessed with the plutonium, there are
significant cost and institutional advantages.

Because of the radiation associated with the *’U, only the
LWBR #**U could be coprocessed with plutonium.

Stand-alone options capable of processing all U
(including ***U with high **U levels) require modification
of the DWPF (HLW glass plant) to handle radiation levels
from “empty” HLW canisters with *’U cans.

The option is potentially viable for the LWBR **U if the
timing issues between the ***U and plutonium program can
be resolved.

 
88

ORNLDWG98C-438R

Dry Blend DUO,

233y0, * r CeramicAdditive

 

 

 

 

 

 

Isostatic Press And Sinter
Ceramic Powder

 

 

 

   

i

\‘;.\\\
Emplace Can in HLW Canister and Add Molten HLW Glass

Molten HLWGIlass

 

Canwith 2°3U Ceramic

  
 
  

 

Disposal

YM-type Geological
Repository

 

Fig.4.13.Can-in-canister 23Udisposition option.
 

 

 

 

 

K

 

 

 

 

  

|
\

|

 

 

 

 

 

 

89

ORNLDWG 97C-39R2

Cans of Ceramic
with Fissile Material
(3.5x151n.)

™ Metal Frame to Hold

Cans in Position

v DWPF
High-Level-Waste
Canister ( 2 x 10ft)

\

High-Level
Waste

 

 

 

 

 

 

Fig. 4.14.Can-in-canister fordisposition of 2*U orplutonium.
90

From the perspective of the plutonium immobilization program, there are advantages and
disadvantages. The thorium would be expected to (1) increase the difficulty of recovery of the
plutonium from the ceramic and (2) improve waste form performance because ThO, is significantly
more inert than UO,. The replacement of the depleted UO, would reduce the available DU for
long-term criticality control in the repository.

* Separate processing. With separate processing in the proposed plutonium immobilization facility,
processing times would be significantly increased because the 14 tons of ThO, in the LWBR ***U
1s an added feed to be processed, not replacements for depleted UO,. With separate processing, a
custom ceramic would need to be developed and qualified for this material. The quantity of waste
to the repository would increase by at least one order of magnitude.

There is the option to create a stand-alone can-in-canister facility for *U. The stand-alone option
would be a remote-operated facility which would allow processing of any ***U with any level of **U and
any level of radiation. Such a facility would be similar to the previously described aqueous (Sect. 4.3) or
dry-blend (Sect.4.5) options. The uranium would be isotopically diluted with DU and converted to an inert
ceramic. Potential waste forms include UO, or a mixture of UO, and ThO,. The addition of ThO, would
increase processing temperatures but produce a better waste form. The product would be put into minicans
that would than be placed into empty HLW canisters. HLW glass would be poured around the minicans.
The primary advantage of this option would be to create an HLW (can-in-canister) glass form similar to
that already accepted by the repository (Paperiello January 25, 1999).

There are complications with a stand-alone can-in-canister option in addition to the need for a separate
RH processing facility. Significant modifications may be required to the DWPF—the HLLW vitrification
plant at SRS. The DWPF was designed on the assumption that the “empty” HLW canisters had low
radiation levels and could be manually prepared and fed into the plant. HLW canisters containing
plutonium ceramic cans but no HLW glass have low radiation levels and match this facilities design basis.
None of the areas on the cold-side of the plant that handle empty HLW canisters have appropriate radiation
shielding for canisters with **U containing significant quantities of **?U and associated high-gamma
radiation fields. These areas of the plant would have to be modified to allow transfer of canisters
containing cans of ***U into the existing high-radiation areas in the facility.

The previous considerations indicate that the can-in-canister option is a potentially viable for the

LWBR ***U; but, there are major difficulties in processing other ***U in the inventory.

4.8 RH TRUW PROCESSING (DISPOSAL)

Uranium-233 could be co-processed with DU and RH TRUW to produce a waste form acceptable for
existing *U waste storage facilities or a WIPP-type disposal facility. If the appropriate processing

technology is chosen, the **U would be isotopically diluted with DU. Because the handling requirements
91

for RH TRUW are similar to those required to handle ***U, RH TRUW processing facilities would be
expected to be able to process ***U with minimal modifications. The ***U has many of the characteristics of
RH TRUW, and, historically, DOE has treated wastes with significant quantities of **U as TRUW and
much of this waste as RH TRU waste. A summary of the option is shown in Table 4.13.

Currently, no RH TRU waste processing facilities exist; however, DOE plans to build such facilities to
process and package RH TRU into forms acceptable for WIPP. Most of the inventory of RH TRU waste
(DOE December 1996) generated in the United States is at two sites, ORNL and Hanford. Most of the
future RH TRU waste will also be generated at these sites. Future facilities must ultimately be built at
these sites. The final waste volumes would depend upon the specific processes that were chosen for

processing and packaging RH TRU wastes.
4.8.1 ORNL

ORNL is considering several options for newly generated RH TRUW, including a small vitrification
system for liquid wastes. Much of the newly generated RH TRUW at ORNL is liquid waste. If a small
vitrification facility is built, the ***U could be added to this waste stream with DU and co-processed with
future RH TRU wastes (Fig. 4.15).

One specific variant of this option deserves special note: uranium can form part of the glass structure.
This creates the possibility of using the uranium (DU and ***U) as a glass former to incorporate other
wastes into the glass. If this can be accomplished, the volume of the waste taken up by the ***U would be
reduced. The uranium would make up part of the waste form, not just the waste. However, the small,
annual RH TRU waste generation rate would lead to a protracted ***U disposal scenario. There are two
uncertainties associated with this variant. The first uncertainty is the allowable loading of uranium in
glass. Low waste loadings imply high waste volumes and correspondingly high transport and disposal
costs. The second uncertainty is whether any facilities will be built that are compatible with disposition of

233U. No decision date for treatment of these wastes has been announced.
4.8.2 Hanford

DOE’s Hanford site in Washington state has a large inventory of solid RH TRUWSs (no liquids). As
part of the Triparty Agreement between DOE, EPA, and the State of Washington, a study will be initiated
on methods to process this waste into a form acceptable for shipment to WIPP. The study has not yet been
mitiated. It is to be completed by June 1999 as Milestone 99 of the Triparty Agreement. The viability of
using this facility for processing ***U depends upon both a decision to build the facility and the technology
chosen. If a technology such as plasma-torch processing is chosen for volume reduction, it would be
applicable to **U disposition. If the facility is only a repackaging facility, it would not be suitable for a

¥V disposition mission.
92

Table 4.13. Summary: Co-processing >**U with RH TRUW forms

 

Application
Acceptable **U feeds

Method to convert to non-weapons-usable
233
U

Disposal site
Criticality control strategy

Technical description

Assessment
Advantages

Disadvantages

Evaluation

Disposal
All

Isotopic dilution

WIPP-type repository
Isotopic dilution

U-233 is coprocessed with proposed RH TRUW
processing facilities. Process depends upon the specific
facility.

Low cost

The option depends upon future construction of RH
TRUW processing facilities at either ORNL or Hanford.
No decisions have been made as to whether such facilities
should be built.

The option exists only if RH TRUW processing facilities
are built and the technology for such facilities is suitable
for *°U isotopic dilution with DU. No decisions are
expected in the near-term.

 
93

ORNLDWG98C-431R2

233 U

 

DEPLETED
URANIUM O XIDE

 

NITRICACID

 

 

 

 

 

 

 

DISSOLUTION
ANDMIXING

RHLIQUIDTRANSURANICWASTE

 

 

 

 

 

 

 

 

 

WIPP-TYPE
GEOLOGICAL
REPOSITORY
TRU
GLASS

 

Fig.4.15.Co-processingof liquidRH TRUW and Z*Ufor dispositionin a WIPP-type repository.
94
4.9 CH TRUW PROCESSING (DISPOSAL)

The United States is considering the building of one or more CH TRUW processing facilities to
(1) convert untreated CH TRUW into treated CH TRUW suitable for disposal at WIPP and (2) reduce the
volume and cost of waste transport and disposal. Some of the potential treatment technologies would allow
simultaneous feeding of **U, DU, and CH TRUW to the process with isotopic mixing of the ***U and DU
within the process. Not all treatment process are capable of achieving this objective. As a consequence,
the viability of this option depends upon (1) a decision by DOE to build a CH TRUW treatment facility
and (2) selection of a technology that assures isotopic mixing of **U and DU. Table 4.14 summarizes this
option.

A CH TRUW treatment facility is being built at INEEL; however, the processes chosen for that facility
are not capable of isotopically mixing **’U with DU (see Sect. 4.6, “Waste Threshold Options™). Decisions
have yet to be made at a number of sites on whether to build CH TRUW treatment facilities.

An example of a CH TRUW treatment process that would be usable for disposition of ***U is the
plasma torch process (Fig. 4.16). Pilot-plant experiments (S. D. Poling et al. March 9, 1994; R. L. Gillins
and S. D. Poling, May 10, 1994; G. R. Hassel, R. M. Geimer, J. A. Batdorf, and G. L. Leatherman,

May 10, 1994) have been conducted on this technology for CH TRUW treatment. In this process, the
wastes are heated by a plasma torch to sufficiently high temperatures that organics are pyrolyzed and other
wastes are converted to a glassy slag. A plasma torch heats the wastes. After the organics have been
oxidized or pyrolyzed and a slag has formed, the slag flows through a hole at the bottom of the hearth into
a waste container.

Isotopic dilution is assured by codissolution of the %*U and DU into the slag. The slag provides a good
matrix to contain the *’U. The dense slag also provides significant radiation shielding against ***U decay
product gamma radiation. There are several unresolved issues for this and other candidate process options:

WIPP acceptance. The plasma torch process can accept CH TRUW, but not high concentrations

of uranium. Too much uranium would raise the melting point of the slag until the molten slag
could not flow into the WP. There are similar technical constraints with most other CH TRUW
treatment options. This implies that such options would require spreading the ***U over a large
number of WPs that also contain TRUW. Because of the large number of packages, it would not
be desirable to add this material to existing >**U waste storage areas. It should be directly shipped
to WIPP. WIPP is authorized to accept TRUW and TRUW containing **U, but not ***U wastes.

In this case, the processing option mixes two previously separate streams. The regulatory status of
such processed wastes must be clarified.
95

Table 4.14. Summary: Convert ***U to CH TRUW form“

 

Application
Acceptable **U feeds

Method to convert to non-weapons-usable
233
U

Disposal site
Criticality control strategy

Technical description

Assessment
Advantages

Disadvantages

Evaluation

Disposal
LWBR, All?

Isotopic dilution, chemical dilution

WIPP-type repository
Isotopic dilution

Excess ~°U, DU, and CH TRUW are coprocessed into a
processed CH final waste form.

Potentially a low-cost option

The option may increase radiation levels associated with
CH TRUW. Radiation exposures probably limit option to
LWBR #**U unless other ***U is processed with large
volumes of CH TRUW.

The option co-mixes ***U and existing CH TRUW. There
are uncertainties in whether such material can be accepted
by WIPP.

The option exists only if CH TRUW processing facilities
are built and the technology for such facilities is suitable
for *U isotopic dilution with DU.

No decisions are expected in the near-term.

 

“This option, when applied to LWBR **’U, can be considered a variant of the waste threshold option.
96

ORNLDWG98C-498R4

Plasma-Torch or
Equivalent Isotopic Dilution Process

 

Plasma Torch
DU /

233 CH TRUW

Molten Slag

    
 

 

   
 

   

     

N

3
PR ‘\\fi.\ 230 "b&l‘}:’ e
A e
%\&\:&\ \\\‘@\ S i
=5

 

              
   

it el ‘W\&g\

S N

*“%%\‘3\}*““ IR e
Mo - \Q\ ‘:’\'S{; \W}\ 2

R
g = e
; L

      

      

 

 

 

 

2
N R
\\\\}\x\\ \\\Q\@\m\\\\. T
= e
\{-\\ % fi\\&i‘:\‘ el
2 \\::\\i\\: \\\R\%\:&\\\;@\\\
\\; t\\:‘:‘:\- A
s - e SRR iR
BAARANS

233U Waste WIPP-type
Storage Area Facility

Fig.4.16.Uranium-233 dispositionbyco-processingwith CHTRUW.
97

* Radiation exposure. Radiation levels are limited by dilution and self shielding. However, this
option may increase total occupational exposure. The potential problems with radiation exposure
are strongly dependent upon the ***U feed material. The LWBR ***U has the lowest radiation
levels, and, thus, would be the most acceptable feed.

4.10 LWR FRESH FUEL (DISPOSAL)
4.10.1 General Description

One option for the disposition of ***U is to use it in the fabrication of LWR fuel and its subsequent
irradiation in a commercial nuclear reactor. To effect this option, the surplus **U would be blended with
either DU or enriched, commercial-grade uranium to produce suitable fuel material. Table 4.15
summarizes the option.

The blending of ***U-containing materials with DU (or enriched, commercial-grade uranium) is unique
because of the unusual isotopes involved (most notably, **U and **U). The relatively small quantities of
*¥U (as compared to the U.S. HEU inventory) would result in an expensive, special campaign. However,
some of the ***U materials are similar to HEU (i.e., they contain primarily **U). As a result, it is
reasonable to consider the inclusion of these materials in a special campaign that is being planned for some
HEU materials.

One specific option examined in this section is the blending of the surplus U with DU and the
surplus, off-specification HEU in storage at SRS, INEEL, and Y-12. This off-specification HEU has a
high isotopic content of ***U and **°U (McWhorter 1995). Once irradiated, fuel containing the surplus ***U
would be destined for permanent disposal in the SNF repository.

The down-blending of HEU with DU results in a material with characteristics different from those of
LEU produced directly through the enrichment process. Most notably, this down-blended material will
have a higher isotopic content of **U than that of the LEU produced directly from enrichment.
Additionally, the surplus, off-specification HEU has a high U content which will in turn results in a high
2% content in the LEU that is produced from this material. As a result, a special campaign is being
planned to down-blend the off-specification HEU to LEU.

Three different scenarios for the blending of the surplus **U with off-specification HEU have been
analyzed: (1) disposition of ***U that is in storage at the Y-12 facility in Oak Ridge, Tennessee;

(2) disposition of the CEUSP ***U; and (3) disposition of the entire ***U inventory (i.e., all **U-containing
materials). The **U in storage at Y-12 is primarily HEU and is similar to the material being considered for
down-blending at SRS. The scenario for the CEUSP ***U provides an analysis of the worst (in terms of
32U content) material in the ***U inventory. Finally, the scenario for the entire ***U inventory provides an
upper-bound analysis for disposition of U with off-specification HEU. The results from these analyses

are summarized in the following subsections. A schematic of the general option is shown in Fig. 4.17.
98

Table 4.15. Summary: Convert >**U to LWR fuel

 

Application
Acceptable **U feeds

Method to convert to non-weapons-usable
233
U

Disposal site
Criticality control strategy

Technical description

Assessment

Advantages

Disadvantages

Evaluation

Disposal
Y-12 #¥U

Isotopic dilution, radiation

YM-type repository
Isotopic dilution

Excess “U is dissolved in acid; purified; isotopically
diluted with DU, natural or LEU; converted into UO,
pellets; fabricated into fuel assemblies; and sent to power
reactors as fresh fuel. The SNF is sent to the YMR.

Clear path to final repository disposal

Medical isotopes can be recovered during the purification
process.

Except for the *°U stored at the Y-12 plant, this is an
expensive option because existing fuel fabrication plants
can not accept “*U due to the radiation levels.

The Y-12 #*U inventory consists of very pure **°U diluted
with high-grade HEU. The radiation levels of this material
are sufficiently low that it could be fabricated into fuel in
existing facilities.

This is a potentially attractive disposition option for one
small lot of **U which could be processed in existing
facilities. It is an expensive option for the remaining **U
because special fuel fabrication facilities would be
required.

 
99

ORNLDWG 97C-409R
___ Off-Specification

HEU (Optional)
333 U

g

 

     

> |

 

 

 

 

 

 

 

 

 

 

 

 

   

 

DU — Purification (Solvent
Nitric __ Extraction o rlon-Exchange)
Acid

. .
Precipitate
Uranium
Dissolution -
and Mixing \\\\\
. Sinter Press
\\\\\\\\\\\\\\\\\\\\\\\%
\=N=N=N Conversion
to UO, Pellets

 

 

 

R

 

 

  

 

   

Irradiation in a Nuclear Reactor

Fig.4.17.Conversionof 23U t o LWR fuel,irradiation, and disposaloftheSNF inYM.
100
4.10.2 Past Experience

The United States has experience in the manufacture and irradiation of ***U-based fuels. The
Shippingport Atomic Power Station, located near Beaver, Pennsylvania, was the first large-scale central-
station nuclear power plant in the United States. It was started up in 1957, was shut down in 1982, and
was decommissioned over the ensuing few years (DOE July 1992). Shippingport was initially an enriched
*¥U-fueled pressurized LWR, but it was converted in 1977 to an LWBR based on the ***U-thorium fuel
cycle.

To minimize the radiological dose to workers during Shippingport fuel fabrication, the ***U nitrate
solution, processed at ORNL, was purified by ion exchange to remove the ***U decay products and to
provide a window of time during which the radiation field from the material produced was significantly
reduced. The resulting solution was then converted to oxide and shipped, not more than 1 week to 10 d
following purification, to Bettis Atomic Power Laboratory for fuel fabrication in a relatively low radiation
field. It is noteworthy that the largest radiation exposure of personnel under the Naval Reactors Program
occurred in 1975. The combined cumulative dose to all navy program personnel was 15 times greater than
the amount reported in 1994—primarily because of the fabrication of the LWBR core for Shippingport
(Mangeno and Burrows March 1995). Some of the ***U in inventory today is of better quality than the
Shippingport material; however, a significant quantity of the current inventory has higher ***U impurities
and higher radiation levels than the material that was processed. The experience gained from handling ***U

materials indicates that control of radiation doses is the major issue.
4.10.3 Conversion of Off-Specification HEU and **°U to LWR Fuel

DOE and the Tennessee Valley Authority (TVA) have signed a Memorandum of Understanding
(MOU) regarding the conversion of SRS off-specification HEU to commercial nuclear reactor fuel. This
MOU could lead to disposition of as much as 30 t of off-specification HEU (Nuclear Fuel
February 10, 1997; McWhorter April 29, 1997). The HEU will be diluted with natural uranium in order to
meet TVA LEU specifications.

It 1s anticipated that test fuel assemblies containing the down-blended, off-specification HEU will be
fabricated by the spring of 1999. These assemblies will be irradiated for about 6 months in a reactor and
then removed for testing. The results of these tests will be used to obtain permission from the NRC to use

the down-blended, off-specification HEU in TVA nuclear reactors (McWhorter April 15, 1997).
101

The composition of the SRS HEU and that of the materials for the three scenarios studied are shown in
Table 4.16. The SRS DU is ~0.27 wt % ***U and ~99.73 wt % ***U (McWhorter April 29, 1997). The
SRS HEU will be down-blended with DU to obtain ~4.9 wt % LEU. It is estimated that about 108 t of
SRS DU will be required. The resulting composition of the SRS LEU is presented in Table 4.17. Also
presented in Table 4.17 are the compositions for the three proposed blends of materials containing **°U
(i.e., Y-12 U, CEUSP ***U, and all ***U-containing materials) and the SRS LEU. For blending the Y-12
U, no additional DU would be required. However, to maintain an enrichment of 4.9 wt %, for the
examples of the CEUSP U and all #*U-containing materials, about 19.7 and 44.5 t of SRS DU would
have to be added, respectively. In each of the scenarios analyzed, the fissile isotopes **U and **°U are
blended with the SRS LEU. Therefore, the enrichment is defined in terms of the equivalent ***U enrichment

and 1s given by the following formula:

massof U + (%) (massof 23°U)

 

Equivalent **°U enrichment = x 100% . )

totalmassof U

This formula provides an estimate of the equivalent mass of U for the ***U present.
In Table 4.18, the proposed blends are compared with American Society of Testing and Materials
(ASTM) specifications. These specifications are industry standards used by LWR fuel fabricators and are

used to address three primary issues: (1) neutronics, (2) alpha activity, and (3) gamma activity.
4.10.3.1 Neutronics

Uranium-236 is a neutron poison, and the basis for its specification is rooted in neutronics
considerations (Cagle June 25, 1997). The presence of “*°U in a fuel (above the specification) would
require a special analysis of the fuel performance. For each of the scenarios examined, the ***U content
would be well above the specification. However, this off-specification condition is already being addressed
by the SRS disposition program by its fabrication and testing of fuel assemblies for NRC approval.
Additionally, the concentration of ***U in the resulting LEU could be reduced by further blending the LEU

with enriched, commercial-grade uranium.
102

Table 4.16. Composition of candidate materials for down-blending by mixing with SRS DU

 

 

 

Composition (kg)
Total ***U-containing
Radionuclide SRS HEU“? Y-12 23U CEUSP ¥y materials®

=2y 1.80 x 10™ 4.80 x 10 0.15 0.16

23y 0 0.8 101.1 789.2

24U 108 0.0 0.0 0.0

U 5,445 38.7 796.4 835.2

25y 2,133 0.0 0.0 0.0

U 1314 _3.1 _145.0 _174.9
Total U 9,000 42.6 1,042.6 1,799.5

 

“McWhorter, D. L., April 15, 1997. Westinghouse Savannah River Company, Savannah River Site,
Aiken, S.C., personal communication to A. S. Icenhour, Lockheed Martin Energy Research Corp.,

Oak Ridge National Laboratory, Oak Ridge, Tennessee.

This material also contains ~200 disintegrations per minute plutonium and 11 ppm *Tc.

‘Forsberg, C. W. and A. M. Krichinsky, January 1998. Strategy for the Future Use and Disposition
of Uranium-233: Overview, ORNL/TM-13550, Lockheed Martin Energy Research Corp., Oak Ridge
National Laboratory, Oak Ridge, Tennessee.

“The CEUSP material contains ~287 kg of cadmium.
103

Table 4.17. Composition of proposed blends

 

Composition (kg)

Blended SRS Blended SRS LEU and Blended SRS LEU and Blended SRS LEU and

 

Radionuclide LEU* Y-12 23 CEUSP U 233U-containing materials’

By 1.80 x 10* 1.85 x 10* 0.15 0.16
) 0 0.8 101.1 789.2
B4y 108 108 108 108
35y 5,737 5,7715.7 6,586.6 6,692.4
6y 2,133 2,133 2,133 2,133
3y 109,100 109,103.1 128,891.8 153,654.8

Total U 117,078 117,120.6 137,820.6 163,377.5

 

“Obtained by blending 9,000 kg of SRS HEU with 108,078 kg of SRS DU.

?Obtained by blending the SRS LEU and the 42.6 kg of Y-12 HEU.

‘Obtained by blending the SRS LEU, 1,042.6 kg of CEUSP HEU, and an additional 19,700 kg of SRS DU.

“Obtained by blending the SRS LEU, 1,799.5 kg of **U-containing materials, and an additional 44,500 kg of SRS DU.

Table 4.18. Comparison of proposed blends with ASTM specifications

 

 

Radionuclide Blended Blended SRS LEU  Blended SRS LEU Blended SRS LEU and
or quantity Specification” SRS LEU and Y-12 23U and CEUSP °U  2%U-containing materials
By 0.05 ng/gU 0.002 0.002 1.060 1.002
24y 2,000 pg/g U 922 922 784 661
3y <Swt% 4.90 4.93 4.90 4.90
20 250 pe/e U 18,219 18,212 15,477 13,056
Alpha activity* 3,300 Bg/ke 2.58 x 10° 2.60 x 10° 1.32 x 10° 2.71 % 10°

 

«“Standard Specification for Uranium Hexaflouride Enriched to Less Than 5% >*U,” ASTM Standard C996-96, American
Society for Testing and Materials, West Conshohocken, Pennsylvania, 1996.

’Includes contribution from **U as equivalent *5U.

“The alpha activity presented here is compared with the specification for neptunium and plutonium (3,300 Bqg/kg U). For
purposes of this comparison, the alpha activity is estimated from the amounts of **U, *°U, 24U, U, and their daughters.

“The alpha activity for typical HEU is about 2.38 x 10° Bg/kg U, while it is about 2.68 x 10'> Bq/kg Pu for WGP.
104
4.10.3.2 Alpha Activity

The alpha activity specification is designed for the radiological protection of workers, particularly
during fuel fabrication. High levels of alpha activity would dictate special handling and protective
requirements. The alpha activity specification presented in the ASTM Standard is based on the activity of
neptunium and plutonium in the material. For purposes of comparison, the alpha activity for the scenarios
analyzed is estimated from the amounts of ***U, **U, #%U, **°U, and their decay products. Because the
alpha activity for the blended SRS LEU is significantly higher than the specification for normal LEU fuel
fabrication, this high activity may dictate special handling requirements for the SRS LEU. The addition of
the Y-12 ***U to the SRS LEU increases the alpha activity only slightly, while the alpha activity is
increased by about a factor of 5 with the addition of the CEUSP ***U and about a factor of 10 with the
addition of all **U-containing materials. The alpha activity varies with time, primarily because of the
decay of ***U and the ingrowth of ***U daughters. For a material initially containing only ~*U, the alpha
activity will peak after about 10 years of decay.

The alpha activity specification is for LEU LWR fuel fabrication. It should be noted that the alpha
activity for typical HEU is about 2.38 x 10° Bg/kg U (Duerksen May 19, 1997) and is about
2.68 x 10" Bg/kg Pu for WGP (Albright, Berkhout, and Walker 1997). There is a large experience base at
DOE facilities in handling both HEU and WGP. In addition, there is a large experience base at commercial
(i.e., NRC-licensed) facilities in handling HEU for U.S. Navy reactors and for research reactors. As a
result, the high alpha activity estimated for the scenarios analyzed should not preclude the fabrication of

LWR fuel.
4.10.3.3 Gamma Activity

The gamma activity of a material is a concern with respect to radiological protection of personnel. A
large gamma activity can result in significant radiation fields during all aspects of LWR fuel fabrication
and handling before to loading the fuel into a reactor.

A unique feature of *°U is that it contains ***U. The ***U contaminant in **’U can result in a large
gamma activity and dose. A daughter of ***U, *®TI, emits a 2.6 MeV gamma with a half-life of about
3 min. Similar to the alpha activity, the gamma activity will peak after about 10 years of decay. The
concentration of the U determines the gamma activity and ultimately determines the potential dose to
workers. The #*U content of the proposed blend of SRS LEU and Y-12 HEU would be well under the
specification. However, for the CEUSP and all **U-containing materials blends, the ***U specification

would be exceeded.
105
4.10.4 Conclusions

Of the three scenarios considered, only the Y-12 **U appears to be a viable candidate material for
disposition with the SRS off-specification HEU as LWR fuel. The resulting blend would be almost
indistinguishable from the proposed SRS LEU blend and should have no significant impact on operations
because of its unique characteristics (i.e., it is primarily HEU with a small amount of ***U and #*U). No
legal barrier has been identified that would prevent the blending of the Y-12 *°U with the off-specification
HEU and DU.

For the CEUSP and all **U-containing materials, the ASTM specifications are significantly exceeded
for both **U content and alpha activity. Hence, these materials present a radiological problem for both
gamma and alpha radiation that likely could not be addressed by the current plans for processing the off-
specification HEU to LWR fuel. Before their processing, these materials could be purified by removing the
#2U decay products, thereby reducing radiation levels and thus affording a window of opportunity for
processing the ***U in a lower radiation field. However, because the ***U decay products would build up
rather rapidly, such an option would require close coordination and strict schedules. It is unlikely that it
would be possible to purify the material (by removing ***U decay products), blend the resulting material
with the off-specification HEU and DU, convert the blend to oxide, fabricate pellets and fuel, and install the
fuel assembly in a reactor before the radiation levels increased significantly. Therefore, the CEUSP ***U
and all *U-containing materials are not viable options for blending with the SRS off-specification HEU.
Furthermore, because of the high gamma radioactivity, it is likely (based on earlier experience) that only a

custom-built, remote-operated, fuel-fabrication facility could convert these materials to LWR fuel.
4.11 DEEP BOREHOLE (DISPOSAL)

This section describes the deep borehole concept for geologic disposal of immobilized excess ***U-
bearing material. It is summarized in Table 4.19. The concept is based on a method of direct geologic
disposal. Most of the discussion presented about this concept is based on a documented analysis by
Lawrence Livermore National Laboratory (LLNL) for the disposition of WGP (Wijesinghe et al.
August 23, 1996) and information provided by LLNL staff involved with that analysis
(Halsey August 6, 1998; Wijesinghe August 6, 1998).
106

Table 4.19. Summary: Characteristics of deep borehole disposal

 

Application
Acceptable **U feeds

Method to convert to non-weapons-
usable *°U

Disposal site

Criticality control strategy

Technical description

Assessment

Advantages

Disadvantages

Evaluation

Disposal
All

Immobilization and isolation in deep underground boreholes

Deep underground boreholes in geologically stable rock. (One
borehole may be enough for the disposition of all excess
domestic Z*U-bearing materials.)

Chemical form and physical dilution

Excess ~’U-bearing materials are packaged in canisters and
placed about 2 to 4 km down into an underground borehole dug
into geologically stable rock formation. The emplaced **U
canisters are cemented in place, and then the isolation zone
above the canisters in the borehole shaft is filled with
appropriate backfill materials to provide a significant transport
barrier from the biosphere.

Uses relatively few bulk processing steps
Requires modest transportation requirements

Offers fewer material accounting uncertainties and fewer
opportunities for theft than other alternatives

Provides potentially permanent isolation from the biosphere

One borehole could be enough for the disposition of all >**U-
bearing materials declared excess.

Could expect only small amounts of water migration from a
deep-level emplacement region.

Difficult for such a facility to gain public acceptance and
obtain licensing approval because of perceived political
sensitivities and problems in dealing with another federal
geologic repository that would be following the WIPP and the
YMR.

Difficult to demonstrate subcriticality over geologic time

Technically viable, but with major institutional problems.
Siting and licensing a new waste disposal facility has
historically required a multidecade, multibillion dollar effort.

 
107
4.11.1 General Description

The deep-borehole concept for geologic disposal of excess *?U-bearing material includes the
emplacement of the excess material into the lower part of one or more underground boreholes, each of
which is drilled into a stable rock formation. The stability of the rock formation must be based on
appropriate criteria that cover tectonic, hydrology, thermal, and geochemical characteristics. Application
of deep boreholes as a disposition option for excess WGP was studied extensively by LLNL and is
documented in Alternative Technical Summary Report for Immobilized Disposition in Deep Boreholes,
UCRL-LR-121736 (Wijesinghe et al. August 23, 1996).

If the borehole disposition concept is adopted, >**U-bearing materials would be emplaced and sealed in
a region of stable rock that lies 2 to 4 km below the surface. This region is called the emplacement zone.
A suitable rock identified for borehole emplacement is granite. Following emplacement, the region above
the disposed material, extending from the top of the emplacement zone to the ground surface (and called the
1solation zone), is filled and sealed with appropriate backfill materials. The isolation zone provides a large
transport barrier to any disposed material.

As noted in the LLNL disposition study for WGP (Wijesinghe et al. August 23, 1996), the deep
borehole-emplacement depths can be several thousands of meters greater than those of mined geologic
repositories. At such depths, the groundwater is expected to be relatively stagnant and have temperatures
of 75 to 150°C, pressures of 7,500 to 15,000 psi, and dissolved salt content of up to 40 wt %.

Table 4.19 summarizes the major features of the deep borehole application for the disposition of ***U-

bearing materials. Figure 4.18 shows the general features of the option.
4.11.2 Waste Form and Site Characteristics

Desired characteristics of waste forms suitable for immobilized disposition in deep boreholes include a
stable solid form, high resistance to dissolution by subsurface brines, thermal stability, and compositional
stability over the long periods of time. The host rock formation into which the excess material is buried
should be a plutonic-metamorphic, crystalline rock (e.g., granite) in a tectonically, hydrologically,

thermally, and geochemically stable region (Wijesinghe et al. August 23, 1996).
108

ORNLDWG98C-22

 

 

 

 

il .

L N\E%‘:\\“‘Q’Q‘{{\%
L RN

 

 

\W\\\\\\\\x\\\

 

 

 

 

 

 

 

.
_TaHhS 3
.
Mix with Grout 2km X
Emplace by
Bucket/Pumping

Seal and Backfill

 

 

 

 

 

 

 

 

 

 

 

 

Emplaces Canisters
Seal and Backfill

Fig.4.18.Borehole disposal of 2U.
109

4.11.3 Deep-Borehole Alternative Applications

Two alternatives have been identified for the deep-borehole disposal option (Wijesinghe et. al
August 23, 1996): immobilized disposition and direct disposition. Process flow diagrams for these
alternatives are given in Fig. 4.19. Both alternatives require the transport of excess materials for
disposition to a facility for conversion and repackaging before these materials are transferred to the
borehole facility for emplacement, sealing, and backfill. In the immobilized disposition alternative, an
additional step is included in the conversion phase to immobilize the material before borehole emplacement.
The type of immobilization used would depend on the particular waste form and could include any (or a
combination) of the following processes: grouting, encapsulation in ceramic pellets, or vitrification.
Figure 4.20 illustrates the concept for the immobilized disposal of coated ceramic pellets in grout. For
most of the domestic #*U inventory, little material conversion and some repackaging would be necessary.
Consequently, the appropriate application of the deep borehole option for any excess ***U-bearing materials

would be the direct disposition alternative.
4.11.4 Issues

An evaluation of the deep-borehole option for plutonium disposition was made with other alternatives
and documented by DOE’s Office of Arms Control and Nonproliferation in the report, Final
Nonproliferation and Arms Control Assessment of Weapons-Usable Fissile Material Storage and Excess
Plutonium Disposition Alternatives, (DOE January 1997). Several major advantages and disadvantages of
the deep-borehole option have been identified.

A major advantage of the borehole option is that the isolation zone provides a large transport barrier to
any disposed material. For all practical purposes, this material can be expected to remain permanently
1solated from the biosphere. This feature assumes the availability and selection of a facility that is located
in a geologically stable location which has stagnant groundwater at deep emplacement depths. The
borehole option also uses relatively few processing steps in transforming the material into a form for
disposition. Transportation requirements are also modest. The borehole option also has lower material
accounting uncertainties and fewer opportunities for theft than other alternatives.

Primary disadvantages of the deep-borehole option are related to political sensitivities and public
perception. The deep-borehole option requires the availability and selection of a facility that is located in a
very geologically stable location that has stagnant groundwater at deep emplacement depths. It may be
difficult for a deep borehole facility to acquire public acceptance and obtain licensing approval because of
problems in these areas experienced from licensing geologic repositories like WIPP and the YMR.

A borehole facility for the disposition of ***U-bearing materials would be another federal waste geologic
disposal facility that would require public review and approval for licensing. Another problem with the
deep borehole option involves criticality verification. In the deep-borehole environment, it would be

difficult to demonstrate or verify subcriticality of the disposed materials over geologic time.
ORNL 98-6911A/arb

 

 

Pits Deep Borehole

Clean Metals Immobilized Disposition Concept

Impure Metals N
Impure Oxide Borehole Facility

Clean Oxide
Alloys
Compounds
Rich Scrap
Miscellaneous
Reactor Fuel

    

 

       
  

    
      
   
   

 

Disassembly, Conversion
& Immobilization Facility

SST Mode SST Mode
AAE At

 
 

 

 

  
 

 

 

 

 

 

 

 

 

Fissile Material — ! .Qn"" Immobilized
— SRS 6M/2R-like Ceramic Pellets
—— Hanford Packages (non-pit) New Package

— Pantex Model FL or Required: 2
[ Fasiy Flas AT400A (pits) Westinghouse ~ ix with Grout

Emplace By
— ANL-W
— INEEL Typa B Drum Bucket / Pumping

— LLNL

 

 

 

 

 

 

 

 

......

Pits Deep Borehole

Clean Metals Direct Disposition Concept

    

 

Impure Metals
Impure Oxide
Clean Oxide

     
   
 
 
 
   
 

  

Disassembly &
Conversion Facility

 

 

 

 

    

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

SST Mode SST Mode
J—I—l\ AR A

Miscellaneous | [ " l IIE
I Fissile Material ! ool | Metal & Oxide
— SRS 6M/2R-like 6M/2R-like
— Hanford Packages (non-pit) Packages
— Pantex Model FL or
— Rocky Flats AT400A (pits) PCV's
— ANL-W ;
— INEEL Canisters
L LANL Emplace Canisters
LN Seal& Backfill

 

 

 

Fig. 4.19. Process flow diagrams for the immobilized and direct deep borehole disposition
alternatives. [Courtesy of LLNL (Wijesinghe et al. August 23, 1996.)]

OTT
111

ORNL 98-6912A/arb

 

TYPICALARRANGEMENT OF DEEP BOREHOLE
WITH EMPLACED CERAMIC PELLET-GROUT MIX

 

SURFACE
FACILITIES

 

“AQUIFER _ T
e T
~ _~ _pomous _ §

SolL

—_— e — _——

 
  
  
 
 
   
   
   

ISOLATION ZONE
SEALANT

— — CASING
HOST ROCK
BOUNDARY GROUT
ISOLATION
ZONE /
GEOLOGIC ENVIRONMENT BACKFILL/
TECTONICALLY, /_ SEALANT
COLUMN

HYDROLOGICALLY,

THERMALLY AND SEALED
GEOCHEMICALLY DAMAGED
STABLE: ROCK ZONE

ANCIENT WATER

 

A 2 KM

UNDERCUT
SEALS

1 UNDERCUT

EMPLACEMENT
ZONE /— PR

CERAMIC PELLET-
GROUT MIX IN
UNCASED HOLE

==

0.44 KM
l » BOTTOM
Y 4 KM ,.\/ s B

EMPIRE STATE
NOTE: HORIZONTAL SCALE EXAGGERATED
BUILDING © ORIZO sC GG

CERAMIC PELLETS
AND GROUT MIX

 

 

 

 

 

 

 

Fig. 4.20. The deep borehole disposal concept for immobilized disposal of coated ceramic pellets in grout.
112
4.11.5 Conclusions

The deep-borehole concept is a technologically viable option. The drilling technology exists, and some
existing sites could be modified and developed to support this option. A major incentive for the deep-
borehole option is that it potentially provides permanent isolation of the disposed material from the
biosphere. However, there are significant hurdles to clear regarding politics, public approval, and licensing
before the deep borehole option could be implemented for **U-bearing materials. It is not a practical

option for the small quantities of **U.
4.12 GREATER CONFINEMENT DISPOSAL (GCD) (DISPOSAL)

The Nuclear Waste Policy Act (1982) and the Nuclear Waste Policy Amendments Act (1987) stipulate
the requirements for disposing of radioactive wastes. HLWs are to be disposed of in a geologic repository
(e.g., the proposed YM site). Many LLWs are suitable for near-surface disposal (e.g., trenches at the
facility in Barnwell, South Carolina). However, some LLLWs are not suitable for near-surface disposal
(i.e., the GTCC or special-case wastes). To provide more isolation of and protection from these wastes,
they may be disposed of at depths that are intermediate between those of near-surface disposal and those of
a geologic repository. GCD was implemented by DOE to provide an intermediate-depth disposal option. It
is no longer used. It may be a potential future disposition option for ’U. Table 4.20 summarizes this
option.

The Nevada Test Site (NTS) has, in the past, operated a GCD disposal facility. As discussed below,
that facility is currently shut down. In addition, the NTS operates several large LLW burial sites. Among
these sites are deep craters from nuclear weapons tests that are being used as disposal sites. Some of these
disposal sites have technical characteristics similar to a GCD disposal facility. The NTS does not
necessarily exclude the disposal of GTCC wastes. In the NTS waste acceptance document (DOE
September 1996), DOE states:

“Disposition of commercial or DOE-generated waste designated as Greater-than-Class C, as defined in

10CFR61.55, may be evaluated for disposal on a case-by-case basis depending on the site-specific

waste classification limits. This review may involve considering non routine disposal operations (i.e.,

controlling depth of disposal, considering other waste forms and package integrity, limiting the other

types of wastes disposed nearby) or the development of a specific radiological performance
assessment.”

Uranium-233 wastes could be classified as GTCC wastes. However, it is noted that the NTS does not

accept TRUW. While *°U is not legally defined as TRUW, it has many of the characteristics of TRUW.
113

Table 4.20. Summary: GCD of U

 

Application
Acceptable **U feeds

Method to convert to non-weapons-usable
233
U

Disposal site
Criticality control strategy

Technical description

Assessment

Advantages

Disadvantages

Evaluation

Disposal
All

Isotopic dilution

GCD facility
Isotopic dilution, chemical dilution

Excess ~°U is isotopically diluted with DU and disposed of
in intermediate depth boreholes. Any one of several
technologies may be used to isotopically dilute *U with
DU.

The construction costs for building a GCD facility should
be less than for a repository.

The single GCD facility (NTS) in the United States has
been shut down by the regulator.

No performance assessment (PA) has been done to
determine the capability of this disposal option for alpha
wastes including **U. Long-term performance of such a
disposal option is unknown. Major development work
would be required to develop the technology and address
performance uncertainties.

Significant resources (time and money) would be required
to determine if this is available option. There is no current
or planned program to develop GCD site technology. A
PA to determine if the single, existing, shutdown facility
will require remediation will not be complete until FY
1999. No currently available data to indicate whether such
a disposal site would be acceptable for 2°U.

 
114

GCD was used at the NTS from 1983 through 1989 to dispose of wastes considered unsuitable for
shallow land burial. The GCD disposal units (Fig.4.21) consist of 36-m deep, 3-m-diam boreholes that
were drilled into the desert alluvium. The boreholes were unlined, except for the top 3 m of each, which
was lined with a corrugated steel culvert. A borehole was filled with WPs to a depth of about 21 m. The
remaining 15 m of the borehole were filled with native soil and a 1.8-m long concrete monument that
indicated the location and contents of the borehole. These boreholes were initially used to dispose of
TRUW, high-specific activity tritium waste, irradiated fuel-rod cladding, and sealed sources. Wastes
disposed of from September 1988 through 1989 contained only *H and DU. Thirteen GCD boreholes have
been prepared. The first borehole, known as the GCD test, was experimental. Seven boreholes were filled
and closed, while three more have been partially filled. Three other boreholes are empty (Shott et al.

June 1995). During operation, there were no specific WAC (as established by a PA) for the facility.
Instead, worker protection requirements (e.g., radiation protection) served as disposal criteria (Ginanni
September 22, 1997; Cochran September 22, 1997).

In 1989, Nevada determined that the boreholes at the GCD facility were injection wells, which are
prohibited by the state. Consequently, DOE suspended operation of the facility and currently has no plans
to resume operations (Ginanni September 22, 1997). In 1994, a preliminary PA of the 4 TRU-containing
boreholes was prepared to evaluate any effects on the public and the environment that could result from
leaving these wastes in place. The final PA for the 4 boreholes is to be issued during FY 1999
(Cochran September 22, 1997).

If GCD were to be considered for 2°U disposition, the technology would have to be fully developed to
meet current waste management requirements. Specifically, a detailed PA is required. The PA is a
methodology to predict the long-term performance of a disposal facility. If the predicted performance does
not match requirements, the design must be altered until the requirements are met. Current information
does not indicate whether a GCD facility such as the earlier facility at Nevada would be acceptable for U
or what the requirements are for a ***U waste form sent to such a facility. Waste management requirements
would likely include acceptable chemical form and require isotopic dilution to avoid criticality issues. In
particular, recent NRC (June 1997) studies suggest the need for isotopic dilution of any ***U sent to a GCD
facility. The NRC analyzed the long-term potential for nuclear criticality at the Envirocare Site in Utah for
disposal of certain radioactive wastes that contained small quantities of HEU. The Envirocare site is a
shallow-land, dry, disposal site that is somewhat similar to the NTS. The results of that analysis indicated
the potential for nuclear criticality due to geochemical concentration of enriched uranium. Isotopic dilution
would be the only viable method for criticality control if such a facility had significant quantities of fissile

uranium isotopes.
115

ORNLDWG98C-21R

 

 

 

 

 

   
  

   
    

 

 

                   

 

3
3
i
N

R

X
N

 

 

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X '{1&{,\

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig. 4.21.GCDoptionsfor3U.

 
116

4.13 SPACE (DISPOSAL)

Disposal of excess **U-bearing materials in outer space is an option that potentially offers the goal of
permanently isolating these materials from the earth’s environment. Table 4.21 and Fig. 4.22 summarize
the option. The possibility of disposal of radioactive waste materials in outer space has been studied
extensively by the National Aeronautics and Space Administration (NASA), DOE, and European sources
(Cosman January 1985).

Several evaluations have been made of space disposal of radioactive wastes, most notably those
documented by Boeing Aerospace Company (BAC) (BAC 1981a, 19815b, 1982a, 1982b) for NASA.

These studies identified and examined all of the major areas associated with the space disposal concept,
including mission and operations analysis and systems integration and evaluation and various options for
space disposal destinations, space transportation, flight support systems, launch sites, launch vehicles, orbit

transfer systems, payload protection, and payload rescue techniques.
4.13.1 General Description

Space disposal of **U-bearing materials, like space disposal for other radioactive waste, would involve
disposal by means of space transportation, not necessarily disposal in outer space itself. Several options
have been identified with this method of disposal, and they are described below.

Disposal of ?*U-bearing materials in space would need to include those activities previously identified
for the space disposal of radioactive materials (Coopersmith August 1992): solidifying the material (if
necessary); embedding the material in an explosion-proof delivery vehicle; and launching that delivery
vehicle into a designated orbit (earth, lunar, or solar). Different orbits are associated with several options

for space disposal.
4.13.2 Waste Form Characteristics

Characteristics of waste forms suitable for space disposal have been identified and discussed in several
sources (BAC 1982bh, McCallum et al. January 1983, and Rice and Priest 1981). Waste forms suitable for
space disposal should be based on chemical and physical compatibly with the required engineered
protection systems. The major space disposal payload requirements identified for candidate radioactive

wastes that also pertain to candidate **’U-bearing materials include:

» Ease of fabrication into a desired form

*  Economics

* High thermal conductivity

» High waste material loading

* Nuclear criticality control

» Resistance to leaching, oxidation, and thermal shock

» Strength of material (i.e., toughness compared with other waste forms)
* Strong thermochemical stability
Table 4.21.

117

Summary: Space disposal

 

Application
Acceptable **U feeds

Method to convert to non-weapons-usable
233
U

Disposal site
Criticality control strategy

Technical description

Assessment

Advantages

Disadvantages

Evaluation

Disposal
All

Isolation in outer space

Outer space (beyond earth’s atmosphere)
Chemical form

Excess ~*U-bearing materials are packaged and placed
into the cargo area of either a booster rocket or space
shuttle for transport and disposition into outer space. Such
materials would need to be solidified (if necessary) and
embedded in an explosion-proof section of a delivery
vehicle, which would be launched for disposition into a
designated orbit (earth, lunar, or solar). The feasibility of
this type of disposition is strongly dependent upon the
technical capabilities of NASA space program.

Provides a mode of permanent disposition that includes
complete isolation from the earth’s environment

Uncertainty in the stability of disposal orbits

Risk of release of material to the atmosphere, either from a
launch failure or a bad orbit, causing a potentially severe
environmental impact with possible international
implications

Restrictive launch characteristics: waste payload system,
launch system and site, and required orbit transfer system

High energy requirements and total costs for space
missions

Viable only if there is significant incentive to use outer
space for the disposition of other types of wastes as well as
*¥U-bearing materials. Very high costs. Significant
environmental, safety, and health (ES&H) questions about
launch safety.

 
118

ORNLDWG98C-482R

Sun Impact

Solar Orbit

High Orbit

        

 

 

 

 

 

 

 

 

 

 

Fig. 4.22.Spacedisposal of 23y,
119

One candidate material form that has been identified for space disposal is cermet, a waste form in
which ceramic particles are uniformly dispersed throughout a metal iron-nickel-copper based matrix. As
discussed in Aaron, Quinby, and Kobisk 1979, advantages of cermet include high material loading, high
thermal conductivity, good corrosion resistance, good leach resistance, high durability, and high mechanical
strength. These features make cermet an attractive shipping form. A comparison of cermet with other

waste forms is provided in Table 4.22 (Rice and Priest 1981).
4.13.3 Space Disposal Options and Destinations
Varied disposal concepts under the space disposal option include:

* Disposal in the Van Allen Belt around the earth by Delta V rocket
* Disposal in high earth orbit by Delta V rocket
» Disposal by Space Shuttle in low earth orbit

For some disposal schemes, several alternative propulsion systems have also been proposed: mass
drivers, solar sailing, and solar-powered thermionic propulsion (Cosman January 1985). One destination
that has been selected in studies (BAC 1981b) as a reference has been the 0.85 astronomical unit

heliocentric orbit destination, which did not require the removal of the protective shield container.
4.13.4 Issues

As described in various sources (notably, Coopersmith August 1992), the benefits of such a disposal
system include the obvious safe, permanent disposal of the material. However, before this objective can be
achieved, several concerns need to be addressed in the consideration of each space disposal option

(Friedlander et al. 1991; Angelo and Buden 1987, and BAC 1982b):

» Stability of disposal or storage orbits

* Long-term risk of reentry of the material into earth’s atmosphere

» Establishment of a supporting technology infrastructure

» Launch characteristics: waste payload system, launch system and site, and required orbit transfer
system

* Rescue mission requirements

Disposal of all radioactive wastes (especially SNF) in space is not practicable because of the high
launch rate required, resulting environmental impacts, high energy requirements, and high costs (Rice and
Priest 1981). A proper selection of a mix of radioactive wastes for space disposal involves consideration of
both the technical and economic factors affected by the energy requirements and total costs of the space
mission(s). Technical factors considered in the waste-mix payload selection affect launch rates and heat

loads. These factors include:
Table 4.22. Potential forms for nuclear waste in space disposal“”

 

Sandia National

 

ORNL ICPP glass Laboratory Borosilicat Metal matrix Hot-pressed
Factor cermet ceramic titanate ceramic e (coated particle) supercalcine
glass
High waste loading M M M L L H
High thermal conductivity H L L L H L
Resistance to thermal shock H H H L H H
Thermochemical stability 1450 1100 1100 1100 1000° 1100
(fabrication temperature), °C
Resistance to leaching H H H H H
Toughness
Applicability to commercial and defense H L L L H L
(Hanford) waste mixes
Fabrication of waste into desired shape M L L H M L
and size
Economics M L
Resistance to oxidation L H H L H

 

“From Rice and Priest 1981.
’Potential is rated as H = high, M = moderate, or L = low.
“‘Copper.
121

» Radionuclide composition and mass content

» Specific activities (and half-lives)

*  Material transport properties

» Potential biological effects

A major issue with space disposal has been its difficulty in addressing concerns regarding a high (or at
least significant) launch risk. The environmental impact of a future launch failure like that of the

1986 Challenger space shuttle disaster is potentially large, depending in large part upon the characteristics

of the payload of radioactive material.
4.13.5 Conclusions

To ensure success, a wide range of technical choices needs to be evaluated along with a combination of
societal, economic, and political factors that must be addressed (Coopersmith August 1992). At best, most
of the concepts considered will not represent viable alternatives to earth-based disposal until well into the
twenty-first century (Cosman January 1985).

Based on the economic and risk assessments made in studies of the space disposal of radioactive wastes
(notably, Rice, Denning, and Friedlander February 1982), in order for the space disposal option to be
viable for **U-bearing materials, it would need to be demonstrated that the overall long-term risks
associated with this activity would be significantly less than the long-term risk associated with any earth
disposal of the current total 2*U inventory. In addition, the long-term risk benefit associated with the space
disposal of **U materials must be achieved within an acceptable short-term and overall program cost.

A major incentive for space disposal (McCallum et al. January 1983) is that it offers a perception of
reduced risks rather than an actual significant reduction. Incremental costs for treating, storing, and
transporting radioactive wastes and ***U materials for space disposal are substantial. A likely space
disposal scenario involving a payload of **?U-bearing materials would be one in which the ***U materials
would be included in the payload along with other forms of radioactive waste.

Cost, schedule, and risk considerations eliminate this option as a viable option at the current time.

4.14 SUBSEABED (DISPOSAL)

Uranium-233 may be disposed of by converting it to a stable form and disposing of it under the ocean
seabed. Placement under the ocean seabed can be accomplished using a drill ship or by tunnelling from

small oceanic islands. Table 4.23 summarizes this option.
122

Table 4.23. Summary: Subseabed disposal

 

Application Disposal
Acceptable **U feeds All

Method to convert to non-weapons-usable  Isotopic dilution and/or geological isolation
233
U

Disposal site Subseabed
Criticality control strategy Isotopic dilution or geology
Technical description Excess **U is converted to a stable waste form and

disposed of under the ocean seabed.
Assessment

Advantages Recovery of final **U waste form would be a lengthy,
costly, and visible to the international community.

Creates the option for an international repository for SNF
with reduced proliferation risks from civilian nuclear
power.

Disadvantages Development of a site is a decade-long, multibillion dollar
effort requiring significant international cooperation and
support.

Evaluation The option is not viable for small quantities of materials
such as ***U because of high development costs, long
development times, and complex regulatory issues that
must be addressed.

 

This disposal option has been examined by multiple international programs. The results of such
studies are that there are major technical (environmental) and long-term institutional (nonproliferation)
advantages of such an option, but significant financial resources and multiple international agreements
would be required to implement such an option. It is, thus, not a viable stand-alone option for the relatively

small quantities of **U. A further description follows.
123
4.14.1 Engineering Description

There are two variants for subseabed disposal (Organization for Economic Co-operation
Development 1988; Forsberg January 1993). The first is direct emplacement of the packaged waste from a
floating offshore platform into the clay layer on the ocean seabed (Fig. 4.23). This option involves the use
of current technologies for offshore drilling of oil. The second option is to construct a repository under a
small oceanic island and under the nearby seabed. This option also includes the possibility of constructing
a man-made island. The technology is well understood because many conventional mines are constructed
under lakes and the ocean by continental seashores. Construction of an ocean island repository would be
no different that construction of a conventional repository on land.

For both variants there are the requirements that sites with stable geology and appropriate geological
formations for waste isolation be selected. There is also a specific set of technical requirements for an
ocean island repository including a groundwater level near sea level, an isolated site, and a site where there
1s significant ocean-water mixing. The basis for these requirements are discussed herein. Large islands
(such as the island of Hawaii) would not meet these requirements and would not be considered ocean-island
repositories because disposal of wastes on islands of such size as Hawaii is similar to disposal of wastes on
continents. Most candidate ocean islands that meet the requirements have land areas of only a few square

kilometers. Many such islands are uninhabited because of the lack of fresh water.
4.14.2 Performance Capabilities

The objective of any radioactive disposal site is to isolate radionuclides from man and the environment
until the radionuclides decay to nonradioactive elements or to such low levels of radioactivity that there is
minimal risk to the public and environment. The expected failure mode of a repository is groundwater-
leaching of the waste, movement of the groundwater to the open environment, and subsequent radioactive
contamination of food or water. Conventional continental geological repositories and subseabed
repositories have many of the same features to ensure waste isolation: (1) insoluble waste forms; (2) long-
lived WPs; (3) deep underground burial [to prevent radionuclide release by natural phenomenon (glaciers,
erosion, etc.) and man-made conditions (terrorists, accidents, etc.)]; (4) radionuclide exchange with local
rock; and (5) geology with low groundwater flow.

The potential attraction of subseabed disposal it that the long-term repository performance may be
significantly better than can be achieved by other disposal options (Fig. 4.24). There is an important
caveat. The most important parameter in determining repository performance is the local site geology. A
good site anywhere can provide high assurances of protection of public health and safety. The differences

in long-term performance are described herein.
124

ORNLDWG98C-424R2

Deep Sea Borehole Island R epository

     
        
    
 
  
 

Unloading
Crane Receiving

Building

 

 

 

  
    
  
     

  

Pipe/CanisterVoid
(grouted)

 

 

........... §§§§§§§&

__________ o

el ——
& A pipeVoid ‘
(grouted)

Sediment/Casing
Void(cemented)

 

   
 

 

 

 

 

   
     
  
 
  
  
  
 

 

 

Package

Backfill

Shaft(300to1500m)

 

\

 

Disposal

" - Tunnel

o / //7_

Underground
Facilities

   

 

 

 

 

Z

 

Bedrock

Fig. 4.23. Subseabed disposal options.
125

ORNL DWG98C-428R2

# Predictability

Hawaii

Midway

Ocean Continent

  
 
      

 

\\ Mid-ocean

   
   

\\ Hotspot

Seafloor Seamounds
Spreading _
(Mid-ocean) Mantle Subzdounc:on

+ Intrinsic Isolation (No Gravity-Driven Water Movement

and Dilution) Rain

   
  
  
  
 

 

 

 

 

 

Oceanic
“Ocean= - -~ T
= —, g Repository
Repository Groundwater Flow
» Waste Form Stability Oxidizing
~—~ Rainwater
Oceanic Continental
S Oxidizing AN
-~ .Ocean= - -
= Groundwater ZZHZ7 o\

 

 
  

(Reducing
Repository Chemistry)

 

Fig.4.24.Characteristicsof seabed disposal options.
126
4.14.2.1 Predictability

The performance of a site is only as good as the capability to reliably predict performance. The ocean
seabed and midocean islands have the most predictable environments on earth. Ocean seabed is formed at
midocean spreading centers by volcanic activity. The seabed moves from these midocean spreading centers
to subduction zones located at the edge of continents. At the edge of the continents, the seabed is subducted
into the earth's crust. The process requires tens of millions of years.

Hot spots under the ocean seabed create volcanic islands such as Hawaii. Because of seabed
spreading, the volcanic cones are carried by the ocean seabed into subduction zones. This results in chains
of islands. For example, in the Hawaiian islands, the island of Hawaii is an active volcano. As one
progresses northwest, the remaining islands are dead volcanos—each older and dead longer than the earlier
island. The dead volcanos slowly sink into the ocean seabed which results in each island becoming smaller
with time. The Hawaiian chain is above the level of the Pacific Ocean until it reaches Midway Island.
Beyond Midway, the dead volcanos are totally submerged and are called the Emperor Seamounds. If one
wants to predict how the island of Hawaii will look in x-million years, one examines an island or seamound
that is x-million years old in the same island chain. The same is true of the oceanic crust.

Because the ocean seabed is under the ocean, changes in climate do not significantly impact subseabed

conditions. This avoids the complications of predicting the effects of climate on a disposal site.
4.14.2.2 Intrinsic Isolation

Subseabed repositories have several mechanisms in addition to those of continental repositories that
isolate radionuclides from the environment.

The ocean reduces the potential for radionuclide leakage from a repository because of geohydrological
factors. The long-term natural mechanism for repository radionuclide release is by groundwater moving
through the repository and then being transported to the open environment. There are two requirements for
water movement: (1) a water-permeable rock (local geology) and (2) a hydraulic gradient to push the water
through the rock from the waste form to the biosphere. The water table of the ocean and a small island is
that of sea level; thus, there is little or no hydraulic gradient to move groundwater and any radionuclides
dissolved in the groundwater. Groundwater moves on the continents because differences in land elevation
above sea level create hydraulic gradients that cause water to flow to the sea. A waste disposal site,
depending upon the local geology, can also create a thermal hydraulic plume from radioactive decay heat;
however, many WPs are designed to last beyond the period of significant radioactive decay heat. The
desirability of avoiding hydraulic gradients (hence, moving groundwater) is the basis for the island-

repository requirement that the island have a groundwater table near sea-level.
127

The potential for human intrusion is significantly lower for island and subseabed sites.

* Resources. Ocean islands and the ocean subseabed are made of basalt—the most common rock on
this planet which covers >50% of the planet’s surface. Mineral deposits and oil are not found in
this type of rock. In both environments, there is no fresh water, hence, no drilling for water.

» Accessibility. Small island repositories and subseabed sites are the most inaccessible locations on
earth. Subseabed sites are accessible only with special deep-ocean capabilities. Small islands have
no fresh water and require significant logistical support to conduct mining operations under such
islands.

4.14.2.3 Independent Isolation Mechanisms

When very-high levels of safety or environmental protection are desired, a facility is designed (1) to
minimize the possibility of failure and (2) to be fail safe. Fail safe, as an engineering concept, refers to
designs that provide for the facility to converge to a safe state if unexpected failures or unfavorable
conditions should occur.

The fail-safe characteristic of an ocean island or subseabed repository is the protective mechanism of
diluting any unplanned releases of radionuclides from the repository by ocean dilution. This is the basis for
the ocean island and subseabed requirement to avoid sites with limited ocean mixing. This environmental
protection mechanism, which is unique to ocean-based concepts, is based on the observations that (1) the
radioactive waste toxicity is small compared with the natural radioactive toxicity of the oceans and (2) very
little of the natural or man-made radioactivity in the ocean reaches man via any pathway, and (3) ocean
dilution usually reduces radiotoxity to biologically insignificant levels.

The advantages of ocean dilution as an independent barrier for a repository have been explicitly noted
by the National Academy of Sciences (December 1983) and British Royal Commissions (September 1976).
The benefits have also been theoretically quantified. The analysis (Cohen 1980) indicates that if
radionuclides were uniformly distributed in the upper continental crust at repository depths, the fraction of
radionuclides that would enter the human food chain each year would be between 10" and 1072, For
radionuclides dumped directly into the ocean, the fraction of radionuclides that would enter the human food
chain each year is between 10 and 10™°. This type of generic analysis indicates that ocean dilution as a
barrier for radionuclide transport to man is within several orders of magnitude as good as continental
geological disposal. The ocean is an excellent, independent, alternative mechanism, backup to either
oceanic island or subseabed disposal.

As an aside, it 1s noted that the natural radioactivity in the ocean exceeds all radioactivity generated by
man. It is estimated that if world electrical demand doubles, all electricity is from nuclear power, nuclear
process is used for 100 y [10° GW (electric)-year of electricity is generated], and all radioactive wastes
were dumped into the ocean, the average radiation dose to marine life would increase by <1% over natural

background radiation levels.
128
4.14.3 Institutional Considerations

There have been many multinational studies of off-shore repositories to address political, economic,
and environmental issues. Such studies have considered such repositories for national use, national
repositories that accept foreign radioactive wastes, and international repositories. There are three major

incentives to operate such a facility as an international repository for all types of radioactive waste:

» Economics. A large repository has economies of scale. It is particularly attractive for small
countries to work together to minimize costs.

»  Environmental. Different countries have different environmental standards. International
repositories are expected to follow international standards that ensure high-levels of environmental
protection.

» Political. An international repository can be used to forward nonproliferation goals by providing a
location for disposal of SNF or fissile materials under appropriate safeguards (Weinberg 1985).
This would be a large-scale implementation of the U.S. policy of accepting return of HEU SNF
that was originally manufactured in the United States as part of our non-proliferation goals.
Return of small quantities of foreign SNF has been controversial. Off-shore repositories offer the
possibility of large-scale international disposal of SNF.

4.14.4 Issues

From technical, economic, and nonproliferation perspectives, there has been a continuous interest in
ocean island and subseabed disposal. However, politically the option is difficult to implement. Several

reasons have been identified for this perspective.

* History. Nuclear weapons testing was conducted in the Pacific by several countries. This resulted
in serious long-term contamination of several islands, deaths of some fishermen, and created a
large political legacy.

» [Institutional. If an international repository is to be built, major international institutions must be
created. This requires a major commitment by national leaders of multiple countries.

» Technical. The technical development requirements for either an ocean island repository or a
subseabed repository are significant. With the ocean island repository option, there are significant
siting constraints. For example, islands such as Bikini (a former test site) might be politically
attractive, but they do not have acceptable geological structures. Other islands that meet the
technical requirements may be considered politically unacceptable. The technical basis has
improved recently with the French decision to declassify geological data on their two Pacific island
underground test sites (International Advisory Committee 1998). These reports provide detailed
data applicable to most Pacific islands.
129
4.15 DISPOSAL AS SNF (DISPOSAL)

The United States is formulating polices for the disposal of research reactor and other miscellaneous
SNF. Much of this fuel contains HEU. Some of this fuel has very low burnup with low associated
radiation fields. The CEUSP and some of the clean **U have higher radiation levels than does this SNF.
The LWBR **°U consists of unirradiated fuel assemblies, fuel rods, and pellets. It has the chemical and
mechanical characteristics of nuclear fuel. These considerations raise the question of whether the material
could be disposed of as SNF. Figure 4.25 and Table 4.24 summarize this option.

SRS is responsible for managing aluminum-clad miscellaneous SNF; INEEL is responsible for
managing other miscellaneous SNF. A draft EIS (DOE December 1998) is being issued on long-term
management options for aluminum-clad SNF—including low-burnup HEU SNF. Options that are being
evaluated include direct disposal, aluminum melt-dilute (see Sect. 4.2.2), and aqueous processing. Some of
these options involve isotopic dilution of the HEU with DU to convert the HEU to non-weapons-usable
LEU. Other options, such as direct disposal, do not require conversion of weapons-usable material to a
non-weapons-usable form. A decision on the preferred management option is expected in FY 1999. If
direct disposal of HEU SNF is chosen, this becomes a potential option for some **U materials. If direct
disposition is not chosen, the option may or may not be viable—depending upon whether a major problem
with this option has been identified. If the option is considered viable for HEU SNF but is not chosen,
costs to develop the option for **U will be much higher since the technology is not being developed for both
HEU SNF and **°U.

DOE is considering direct disposal of the LWBR irradiated SNF. The baseline option for this and
other miscellaneous SNF is codisposal in a repository WP containing five canisters of HLW glass.
Repository criticality safety issues are associated with disposal of this SNF. It is proposed to address these
issues by limiting the amount of SNF per WP by including it in a package containing HLW but no other
SNF (SNF Task Team Report March 1997). In the case of the LWBR SNF, a WP could contain up to
(1) one #*U LWBR blanket SNF assembly with 11.8 kg of **U before irradiation or (2) one ?*U standard
LWBR SNF assembly with 21 kg of **U before irradiation. Detailed repository criticality analysis has not
been done; thus, it is uncertain whether these quantities of material are acceptable. If unirradiated >°U
were to be disposed of, a similar approach would likely be adopted.

For disposal of separated ***U, there are several issues.

» The repository criticality issue has not been addressed for the LWBR SNF; hence, it is uncertain

whether this material is acceptable for the repository. Because this SNF will be in storage for
many decades, there is not an immediate need to address this issue.
130

ORNLDWG 98C-7767

Package **°U

 

  
   

Steel Container

 

 

Place In HLW
Package 5 Each 24-in.
Diameter HLW
77.5-in. OD Glass Canisters

17-in. Diameter
Spent Fuel Canister

 

Repository —

Waste Package

 

-

o

Disposal

\\\\

YM-type Geological
Repository

 

 

Fig.4.25. Co-disposal of 233U(defined as SNF)with HLW.
131

Table 4.24. Summary: Dispose as SNF

 

Application Disposal
Acceptable **U feeds All
Method to convert to non-weapons-usable  Isolation in geological repository
23U
Disposal site YM-type repository
Criticality control strategy Package mass limit
Technical description Excess ~°U is managed as HEU SNF by placing the
material in special packages and then placing the packages
into a WP containing multiple HLW canisters.
Assessment
Advantages Defined path to final repository.
Disadvantages The U does not meet the conventional definition of SNF.
The option may not meet the Spent Fuel Standard
Evaluation Major policy questions must be addressed. It is a low-cost

option if DOE’s SNF program adopts this strategy for
disposal of miscellaneous SNF.

 

e The LWBR **U is not technically SNF. It has not been irradiated in a reactor. The legal status of
treating the material as SNF is unclear. Like plutonium, it is a product of irradiation of targets
with neutrons in nuclear reactors. Any determination on whether plutonium or ***U could be
considered SNF would be applicable to both. This has important implications for U and
plutonium because there are multiple regulatory requirements for managing SNF.

e The *°U may not meet the SNF standard (see Sect. 4.6, “Waste Threshold Option™).

Decisions on disposal of miscellaneous SNF in FY 1999 will likely determine whether this is a viable

option for any “*U. This is an expensive option to develop for a small quantity of material.
132

4.16 SHALLOW-LAND BURIAL (DISPOSAL)

Shallow-land disposal is disposal of wastes in a near-surface structure or trench. It is used to dispose
of low-level radioactive wastes, municipal wastes, and some hazardous wastes. A characteristic of most
LLW that may be disposed of in such facilities is that most LLW becomes nonradioactive in a relatively
short time period. Waste isolation is required only for a limited time period. The quantities of long-lived
radionuclides allowed in a shallow-land disposal site are typically limited to <100 nCi/g.

The radiation characteristics of **U are similar to those of TRU elements such as plutonium. The
primary hazard is from alpha radiation. The United States requires geological disposal of wastes
containing significant quantities of plutonium to minimize the risks from alpha radiation. By definition,
plutonium wastes containing in excess of 100 nCi of plutonium per gram (about 1 ppm) require geological
disposal. If the alpha activity is lower, shallow land disposal as LLLW 1is acceptable. Based on health-
physics considerations, a similar policy would be expected for 2*U. A ***U disposition option based on this

assumption is shown in Table 4.25 and Fig. 4.26.

Table 4.25. Summary: Shallow-land disposal as LLW

 

Application Disposal
Acceptable **U feeds All
Method to convert to non-weapons-usable ~ Chemical dilution
233U
Disposal site Shallow-land disposal
Criticality control strategy Isotopic dilution and chemical dilution
Technical description Excess “*U is diluted by a factor of 100,000 with inert
materials, DU is added, and the mixture is disposed of as
LLW in a shallow-land disposal site.
Assessment
Advantages Meets the Spent Fuel Standard.
Disadvantages High costs are associated with processing and there are
significant uncertainties on how it would be done.
Significant institutional questions on the acceptability of
dilution as a method to change the classification of a waste
form.
Evaluation There are major institutional issues with this option

compared to most other disposition options.

 
133

ORNL DWG98C-7768

Solidification Agent

   
   
 

DU (iluent
LLW Storage
Container
233UO2

 

 

 

Shallow-land
Burial Site

 

 

 

 

 

 

 

 

 

 

 

 

Fig. 4.26.Dispositionof >**Ubydilution toLLW (<100 nCi/g)withshallow-landdisposal.
134

The specific activity of **U is 9.6 x 10° nCi/g. To dilute **U to alpha-radiation levels allowed in
shallow-land disposal facilities, the **U would have to be diluted by about a factor of 100,000. Assuming
that the soil density is 2 g/cm’, the waste volume would be ~50,000 L/kg of *°U.

Diluting **U with large quantities of other materials would be a significant engineering task; however,
the difficulty of large-scale mixing operations and the large quantities of diluent required suggest that the
diluted ***U would meet the Spent Fuel Standard. Based on recent analysis of shallow-land burial grounds
by the NRC (Toran et al. June 1997), DU would need to be added to the waste for criticality control. It
would probably not be necessary to isotopically mix the U with the DU. The different uranium isotopes
could be physically mixed with the solidification agent. The natural mechanisms of uranium dissolution
and transport would be expected to isotopically mix the uranium isotopes over time.

Several factors suggest that this option would be expensive and difficult to implement compared to
other ~°U disposition options.

» Definition of acceptable **U concentrations in LLW. The dividing line on the allowable content
of alpha-emitting radionuclides in LLW to be disposed of in a shallow-land burial ground was
developed for plutonium. Uranium-233 was not considered when this dividing line was
established. Uranium in oxidizing environments, as is often found in shallow-land disposal sites, is
more mobile than plutonium in groundwater. If large quantities of **U were to be disposed of, an
evaluation of the acceptability to dispose of LLW containing <100 nCi/g would be required. There

is a potential that the allowable concentration of ***U acceptable for shallow-land burial would be
lower.

» Dilution as a waste management strategy. Dilution of radioactive waste to change the waste
classification and allow shallow-land disposal is restricted unless there are special conditions. Any
decision on **U could have a large institutional and economic implications for the disposal of
TRUW since there are large quantities of TRUW with somewhat >100 nCi/g of TRU elements.
Consequently, addressing the institutional issues with this option would be complex.

» Economics. The option has high processing and potentially high transport and disposal costs.
This is a direct result of the large final waste volumes. The costs of shallow-land disposal is
strongly dependent on the site. Currently, the NTS accepts wastes with characteristics that are
closest to **U (bomb debris) and that have low disposal costs. For this favorable site, the cost is
$265/m’ ($7.50/ft%). This implies ***U disposal costs in excess of $13 x 10%ton of **U.

4.17 ELECTROMETALLURGICAL PROCESSING (STORAGE AND DISPOSAL)

Uranium-233 may be down-blended with DU using an electrometallurgical process to produce a
uranium metal form for (a) disposal at a YM-type or WIPP-type repository or (b) a long-term storage.

Table 4.26 summarizes the option. Figure 4.27 shows the process
135

Table 4.26. Summary: Electrometallurgical process

 

Application
Acceptable **U feeds

Method to convert to non-weapons-usable
233
U

Disposal site
Criticality control strategy

Technical description

Assessment

Advantages

Disadvantages

Evaluation

Storage, disposal
LWBR, Clean, CEUSP

Isotopic dilution of **U

YM- or WIPP-type repository
Isotopic dilution

33U oxides converted to metal with lithium. **U and DU
metal dissolved in molten salt. Uranium metal
electroplated on cathode as final product for storage and
disposal.

May use partly completed facility at Argonne National
Laboratory (ANL) (West)

It 1s unknown if the product (metallic uranium) can meet
repository WACs.

Complex process

A version of the process is under development to process
SNF where the uranium is recovered for reuse. There are
major technical and institutional uncertainties on whether
the process is viable for *U where the waste form is
uranium metal

 
136

ORNL DWG 99C-109

Reduction to Metal

U0, + + Li

Crucible

 

 

 

  

Furnace\ | |
Uo, +4Li— U + 2Li,O

 

 

Dissolve in Salt/Electrolytic Recovery of Uranium Metal

. / Uranium Cathode

Salt Solution

 

 
 
   
 

Weld Lid

 

    
 

  

WIPP orYM-type
StorageFacility Geological Repository

Fig.4.27.Disposition of 23U by isotopicdilutionwithDU using theelectrometallurgical process.
137

The original process was developed at ANL to recycle uranium and actinides from sodium-bonded,
breeder reactor, SNF into new nuclear fuel. In the electrometallurgical process, metallic SNF is dissolved
in a molten lithium-chloride—potassium-chloride (LiCl-KCl) salt solution. Uranium metal is recovered
from this molten salt by electrolysis for eventual refabrication into fuel for recycling back into the reactor.
The process is currently being investigated for another use—processing certain DOE SNF (National
Research Council 1998) into waste forms acceptable for disposal in a YM-type repository. The uranium
from the SNF is recovered for future use while the fission products and actinides are converted into
acceptable repository waste forms. For this application, the **U is to be down-blended with DU in the
process to produce LEU as a waste or storage form.

For down-blending of ***U, the ***U and DU metal would be dissolved in the molten salt and then the
U and the DU would be codeposited on the cathode in an electrolytic cell as uranium metal. The uranium
metal would be the final product for storage or disposal as a waste form.

The electrometallurgical process is based on the dissolution of the elemental (metal) actinide fuels in
LiCl-KCI molten salt. For those feed materials that are not elemental (metallic) uranium, a front-end
process must be used to reduce the actinide compounds in the fuel material to the elemental state which can
then be dissolved in the molten salt. This requires additional front-end processing for essentially all the
*¥U in the inventory:

e Clean and CEUSP ***U. The clean and CEUSP **U are in the form of uranium oxides. These
forms would be processed using the same methods developed for oxide SNF. Typically, for oxide
fuels, elemental lithium is used as the reductant to convert the uranium oxide to uranium metal.
The oxidized lithium is then converted back to the elemental state for use in the next reduction
cycle. Stainless steel and Zircaloy cladding are not dissolved by this process. The fuel is chopped

so that the fuel meat can be leached from the Zircaloy or stainless steel hulls by the molten salt.
The hulls are then managed as solid, LLW and discarded.

e LWBR?”U. This material is in the form of a high-fired U-Th oxide. It is proposed
(J. Laidler March 5, 1999) that LWBR fuel could be processed by chopping the fuel rods to expose
the pellets, grinding the pellets to submicron size to make it more reactive with the elemental
lithium, reducing the uranium and thorium oxides with lithium metal to the elemental state, and
then dissolving the uranium and thorium in the molten salt mixture.

With current technology, it is estimated that ~1 kg/d of uranium could be processed. Some
laboratory-scale experiments have demonstrated processing “low-fired” U-Th oxide ceramics.
Questions still must be resolved as to whether the process could, in a reasonable time, convert the
“high-fired” ceramics from the LWBR ***U oxides to metal. It has been demonstrated that the
uranium could be separated from the thorium in the electrolytic cell by adjusting the electrolytic-
cell deposition conditions.
138

This process is probably capable of treating the **’U inventory, including the LWBR *’U. However,
there are technical uncertainties that would strongly impact economics. There have been no full system
tests processing high-fired ceramic fuel pellets. The process is complex compared to most other alternative
processes for down-blending ***U.

Moreover, there are significant uncertainties if the uranium product is to be a waste. In all other uses
of this technology, the uranium is a product for reuse. No testing of such a uranium-metal waste form for a
YM-type repository has occurred. It is an unusual waste form. For the CEUSP ***U, which contains
cadmium, it is unclear whether the final product would be acceptable to a YM-type repository which
prohibits chemically hazardous waste forms. It is unknown too, if ***Th (the isotope with potentially
useful medical applications) could be recovered during processing of the clean or CEUSP #*U. Uranium

metal 1s allowed as a storage form although U,Oq is preferred because it is less chemically reactive.
5. CONCLUSIONS

Alternative storage and disposition options have been identified and described. Each option has
different technical, institutional, schedule, and economic characteristics. All of the storage and disposal
options could be implemented; however, the resources (financial, technical, time, and institutional) to

implement the different options vary by orders of magnitude.

5.1 SPECIAL CHARACTERISTICS OF **U
There are several important caveats when considering ***U storage and disposal options.

*  Characteristics. Uranium-233 has fundamentally different characteristics than HEU and
plutonium. With modern ES&H requirements, most of the ***U will require hot-cell processing
because of the buildup of ***U decay products that emit high-energy gamma rays. Storage and
disposition options suitable for HEU or plutonium are not necessarily suitable for *°U.

e Quantities. The quantities of ***U are small compared to HEU and plutonium. In total, there is
<2 t of uranium in somewhat <20 t of material.

e Categories. There are three categories of Z*U with fundamentally different characteristics: clean,
CEUSP, and LWBR **U. Many storage and disposal options are only viable for one or more
categories.

 Institutional constraints. There are a set of unique institutional issues associated with Z*U
disposition options because ***U is the orphan fissile material. Much of the institutional structure
that exists for HEU and plutonium does not exist for >*?U. This includes such items as a lack of an
official definition of weapons-usable *°U to constraints on disposal sites because enabling
legislation did not consider the existence of ***U materials or wastes.

» Criticality control strategy. Criticality control strongly influences costs and feasibility of various
disposition options. In many cases, adding additional DU to minimize nuclear criticality
constraints minimizes costs by minimizing waste volumes.

5.2 STORAGE AND PROCESSING OPTIONS

There are three storage strategies: (1) store as weapons-usable material for future use, (2) store as
non-weapons-usable material for future use, and (3) store for future waste disposal. Weapons-usable *°U
must be stored in a Category I facility to prevent potential theft of material. If the **U is to be stored as
non-weapons-usable material for future use, it must be isotopically diluted to <12 wt % ***U in ***U,
Isotopically diluted material may be stored in facilities with industrial-type security. If the *°U is to be
stored as a waste, it should be isotopically diluted to 0.66 wt % ***U in ***U to convert it to non-weapons-
usable material and minimize the potential for nuclear criticality. This material may be stored in existing

*¥U waste storage areas with ultimate disposal in a WIPP-type geological repository.

139
140

Five technologies have been identified for isotopically diluting ***U with DU. Two processes have been
used on a large-scale: (1) aqueous dissolution and (2) dry blend with sintering. Three processes have been
demonstrated in the laboratory using uranium, but not on an industrial scale: fusion melt, uranium-
aluminum alloy, and electrometallurgical. The borate fusion melt process may be the lowest cost process
to isotopically dilute ***U with DU—if this is the only goal. The uranium-aluminum process can not handle

LWBR ***U because this ***U contains large quantities of ThO, which is incompatible with the process.

5.3 DISPOSAL OPTIONS
Seventeen disposal options were identified. Each of the disposal options has different characteristics:

» The HLW option has the fewest institutional constraints of any option. The disposal costs of the
LWBR **U may be high due to complications from the thorium in this feed material.

* The uranium-aluminum melt blend options depend upon successful development of the technology
for treating aluminum-clad SNF. The option can not process LWBR **U.

» The aqueous process option is the only fully demonstrated process, is the most versatile option
which allows for the recovery of special isotopes, and can produce multiple waste forms that are
acceptable for disposal in a YM- or WIPP-type disposal facility. There are cost uncertainties.

«  The borate-fusion melt option is probably the lowest-cost option for isotopically diluting ***U, but
there are technical uncertainties that could impact costs as well as institutional issues associated
with disposal. It is also a potential pretreatment option to place ***U in HLW tanks.

» The grind, dry-blend, and sinter option is potentially a lower cost option.

«  The waste threshold option is probably the lowest-cost disposal option for the LWBR ***U and a
low-cost option for other **U. There are policy issues associated with this option.

» The can-in-canister option is conceptually identical to the can-in-canister plutonium disposition
option. Because of the radiation levels associated with the *°U, the plutonium facility could only
process the LWBR **°U.

 The RH TRUW processing option is potentially attractive if DOE builds an appropriate RH
TRUW facility. No such facility is planned in the near term.

» The CH TRUW processing option is potentially attractive if DOE builds an appropriate CH
TRUW facility. No such facility is planned in the near term.

e The LWR fuel option for ***U has been demonstrated in the past. Because of the radiation levels of
this feed material, existing fuel fabrication facilities can not process *’U. A new custom-built
facility would be needed. This would include hot-cell handling facilities. It is an expensive option.
There is one exception to this conclusion. The Y-12 **U is HEU with small quantities of high-
purity *’U. It may be convertible into LWR fresh fuel without special facilities.
141

» The borehole-disposal option is identical to that proposed and then dropped for disposal of
plutonium. The #*U variant has the same advantages and disadvantages as the plutonium borehole
option. It is a long-term and expensive option if used to dispose of a single waste.

* The GCD option is highly uncertain. It is unclear whether this option can meet current
environmental requirements for disposal of long-lived radioactive materials.

» The space-disposal option has a very high cost and would require many years to develop. There
are serious questions about rocket-launch safety.

» The subseabed disposal option is environmentally attractive, but it is a long-term and an expensive
option if used to dispose of a single waste.

 Direct disposal of **U as SNF is potentially a low-cost option; but, there are questions whether
this option meets the Spent Fuel Standard. Only the LWBR ***U is a potential candidate for this
disposal option.

« Disposal of *U as LLW (after dilution by a factor of 100,000) has major technical, cost, and
nstitutional barriers.

e Processing and disposal of the **°U electrometallurgical process is complex. It is unclear whether
the final product is acceptable for YM-type repository disposal.
For all disposal options, there are multiple regulatory and other institutional issues that must be
addressed. Those disposal options that coprocess **U with other waste forms have the fewest regulatory
uncertainties. These include HLW glass disposal, the uranium-aluminum melt-dilute options, and can-in-

canister.
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Dry Melt Blending with Natural or Depleted Uranium in the Radiochemical Development Facility
(Building 3019 Complex) at the Oak Ridge National Laboratory (ORNL), ORNL/MD/LTR-24, Rev. 1,
Lockheed Martin Energy Systems, Oak Ridge National Laboratory, Oak Ridge, Tennessee.

Olander, D. L., 1976. Fundamental Aspects of Nuclear Fuel Elements, TID-26711-P1, Energy Research
and Development Administration, Technical Information Center, U.S. Department of Commerce,
Springtield, Virginia.

Organization for Economic Co-operation and Development, Nuclear Energy Agency, 1988. Feasibility of
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Paperiello, Carl J., Director, Office of Nuclear Material Safety and Safeguards, January 25, 1999. Letter
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U.S. Nuclear Regulatory Commission, Washington, D.C.

Peacock, H. B., Jr., T. M. Adams, and N. C. Iyer, September 9, 1998. “Development of the Melt-Dilute
Treatment Technology for Aluminum-Based DOE Spent Nuclear Fuel,” pp. 18-22 in Proc. Third Topical
Meeting on DOE Spent Nuclear Fuel and Fissile Materials Management, Charleston, South Carolina,
September 811, 1998, American Nuclear Society, La Grange Park, Illinois.

Poling, S. D., et al., March 9, 1994. Conceptual Design Report for the Plasma Hearth Process (PHP)
Bench-Scale Radioactive System, SAIC-94/1003, Scientific Applications International Corporation,
Idaho Falls, Idaho.

Rice, E. E. and C. C. Priest, 1981. “An Overview of Nuclear Waste Disposal in Space,” pp. 370-386 in
The Technology of High-Level Nuclear Waste Disposal—Advances in the Science and Engineering of the
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Rice, E. E., R. S. Denning, and A. L. Friedlander, February 1982. Preliminary Risk Assessment for
Nuclear Waste Disposal in Space, Vol. 2, Technical Report, NASA/CR-162029, Battelle Columbus
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Royal Commission on Environmental Pollution, September 1976. Sixth Report, Nuclear Power and the
Environmental Pollution, Her Majesty’s Stationery Office, London.

Savannah River Site, April 1997. Alternative Aluminum Spent Nuclear Fuel Treatment Technology
Development Status Report, WSRC-TR-97-0084, Westinghouse Savannah River Company,
Aiken, South Carolina.
150

Shott, G. J. et al., June 1995. Performance Assessment for the Area 5 Radioactive Waste Management
Site at the Nevada Test Site, Nye County, Nevada, DOE/NV/11432-196, U. S. Department of Energy,
Nevada Operations Office, Las Vegas, Nevada.

SNF Task Team, March 1997. Technical Strategy for Managing of INEEL Spent Nuclear Fuel,
Idaho National Engineering and Environmental Laboratory, Idaho Falls, Idaho.

Toran, L. E., et al., June 1997. The Potential for Criticality Following Disposal of Uranium at Low-
Level Waste Facilities, NUREG/CR-6505, U.S. Nuclear Regulatory Commission, Washington, D.C.

U.S. Congress, October 30, 1992, Waste Isolation Pilot Plant Land Withdrawal Act, Public Law 102-579,
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U.S. Congress, 1996. Waste Isolation Pilot Plant Land Withdrawal Act, Public Law 102-579, as
amended.

U.S. Department of Energy, July 1992. Characteristics of Potential Repository Wastes,
DOE/RW-0184-R1, Vol. 2, Washington, D.C.

U.S. Department of Energy, August 30, 1994. Waste Form Requirements for the Potential Disposition of
Weapons-Usable Fissile Materials in a Deep Geological Repository, AO0O000000-00811-1708-00004,
Rev. 00C, Washington, D.C.

U.S. Department of Energy, March 29, 1995. Summary Report of the Screening Process,
DOE/MD-0002, Washington, D.C.

U.S. Department of Energy, October 1995. Final Environmental Impact Statement: Interim Management
of Nuclear Materials, DOE/EIS-0220, U. S. Department of Energy, Savannah River Site,
Aiken, South Carolina.

U.S. Department of Energy, December 12, 1995. Record of Decision, Notice of Narrowing of
Alternatives, and Notice of Preferred Alternatives, U. S. Department of Energy, Savannah River
Operations Office, Aiken, South Carolina.

U.S. Department of Energy, February 8, 1996, Record of Decision: Materials Stabilization,
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U.S. Department of Energy, April 1996. Waste Acceptance Criteria for the Waste Isolation Pilot Plant,
DOE/WIPP-069, Rev. 5, Carlsbad, New Mexico.

U.S. Department of Energy, Office of Fissile Materials Disposition, June 1996a. Disposition of Surplus
Highly-Enriched Uranium Final Environmental Impact Statement, DOE/EIS-0240, Washington, D.C.

U.S. Department of Energy, June 1996b. Technical Strategy for the Treatment, Packaging, and Disposal
of Aluminum-Based Spent Nuclear Fuel, Washington, D.C.
151

U.S. Department of Energy, July 29, 1996. Record of Decision for the Disposition of Surplus Highly-
Enriched Uranium Final Environmental Impact Statement, Washington, D.C.

U.S. Department of Energy, Nevada Operations Office, September 1996. Nevada Test Site Waste
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U.S. Department of Energy, January 1997. Final Nonproliferation and Arms Control Assessment of
Weapons-Usable Fissile Material Storage and Excess Plutonium Disposition Alternatives,
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U.S. Department of Energy, Carlsbad Area Office, September 1997. Waste Isolation Pilot Plant Disposal
Phase Final Supplemental Environmental Impact Statement, DOE/EIS-0026-S-2, Carlsbad, New Mexico.
U.S. Department of Energy, September 25, 1997. Implementation Plan for Defense Nuclear Facilities
Safety Board Recommendation 97-1: Safe Storage of Uranium-233, U. S. Department of Energy,
Washington, D.C.

U.S. Department of Energy, Idaho Operations Office, November 1997. Plutonium Focus Area Technical
Advisory Panel Recommendation: The United States Should Adopt a Single Policy for the Safeguards
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U.S. Department of Energy, August 1998. Final Environmental Impact Statement on Management of
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U.S. Department of Energy, December 1998a. Report on the Savannah River Site Aluminum-Based Spent
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U.S. Department of Energy Office of Fissile Materials Management, 1999. Summary Report of the
Screening Process to Determine Reasonable Alternatives for Disposition of Uranium-233,
DOE/MD-0012, Washington, D.C.
152

U.S. Department of Energy, Office of Environmental Management, January 1999. Advanced Mixed Waste
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Methodologies for Dissolution and Corrosion of AI-SNF,” pp. 516-524 in Proc. Third Topical Meeting on
DOE Spent Nuclear Fuel and Fissile Materials Management, Charleston, South Carolina,

September 8—11, 1998, American Nuclear Society, La Grange Park, Illinois.

Wijesinghe, A. M., et al., August 23, 1996. Alternative Technical Summary Report for Immobilized
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Wijesinghe, A. M., August 6, 1998. Lawrence Livermore National Laboratory, Livermore, California,
“Re: Request for Borehole Evaluation Report,” correspondence to Stephen N. Storch, Oak Ridge National
Laboratory, Oak Ridge, Tennessee.
Appendix A

RADIATION LEVELS FROM **U
The radiation levels from ***U determine (1) many of the facility requirements for its processing, (2) its
transport requirements, and (3) the waste-handling requirements of any ***U product. For example, if the
¥V is classified as TRUW, it may be either CH or RH waste depending upon the external radiation levels.
For TRUW and most other wastes, the dividing line between CH and RH is 200 mrem/h at the surface of
the container. The radiation levels from **U, as discussed in Sect. 2, depend upon (1) the impurities in
23U, particularly 22U and its decay products; (2) the age of the *’U since the gamma-emitting decay
products have been removed from the ***U and ***U impurity; and (3) other materials that the *°U is mixed
with. This appendix provides some additional information on expected radiation levels.

A set of calculations was made to determine the radiation dose from a 55-gal (208-L) drum near the
surface of the drum. Radiation doses were calculated 1 cm from the drum as an approximation for surface
measurements on the drum (minimize numerical instabilities in radiation calculations). The following

assumptions were used.

o Uranium-233 impurity level. The *’U was assumed to have 100 ppm **U. The **U inventory
has materials with ***U concentrations from a few ppm to somewhat >100 ppm.

e Uranium-232 age. The primary radiation from **’U in storage is from the ***U decay product,
*8T]. This decay product emits a 2.6-MeV gamma ray. If the uranium has been purified, the **TI
builds up over time and then decays as the ***U decays. The time of maximum radiation levels is
10.3 years after separation of the decay products from the uranium. The radiation calculations
herein are for this particular time of maximum radiation.

*  Drum characteristics. The drum height is 35 in., the diameter is 24 in., and the wall thickness 1is
1/16-1n. carbon steel.

*  Uranium chemical form. The uranium is assumed to be U;Oq in the form of a loose powder with a
density of 1.5 g/cm’. The drum contains ~390 kg of oxide.

If the **°U is isotopically diluted with DU to become non-weapons-usable **U (1 part ***U with
7.407 parts DU containing 0.2 wt % ***U), the external radiation doses calculated 1 cm from the drum will
be 141 R/h. The ***U concentration would have to be <0.1 ppm to be CH material (<200 mrem/h). In this
specific example, a neutron absorber would have to be added to the drum for criticality control.

If the **°U is isotopically diluted with DU to minimize criticality concerns (1 part 2*U with 188 parts
DU containing 0.2 wt % **°U), the external radiation doses calculated 1 cm from the drum will be
6.247 R/h. The #**U concentration would have to be reduced to <3 ppm to reduce the radiation levels to
those of CH waste (<200 mrem/h). This implies that a CH drum (without special features) can contain up
to a few tens of grams of **U with high concentrations of **U (>100 ppm ***U) or a few hundred grams of

3 with lower 2*U concentrations (105 ppm *2U).

A-3
Appendix B

RECOVERY OF **Th FROM ***U FOR MEDICAL USE

B-1
B.1 MEDICAL USES OF BISMUTH-213

One potential large-scale use of ***U involves one of its decay products, ?’Bi (Bereolos et al.

June 1998). Over the past decade, there has been considerable research in the area of radioimmunotherapy
using alpha receptors. Specifically of interest are antitumor antibodies radiolabeled with an alpha emitter
(Knapp and Mirzadeh 1994; Geerlings 1993). In this method, the isotopes are attached to antibodies that
specifically target the cancer cell, after which the resulting alpha emissions kill these cells with high
efficiency.

Previous work in this area focused on using *'*Bi, which is produced by the decay chain of ***U (or
*Th). However, the undesirable side effect of *'*Bi is the 2.6-MeV gamma radiation emitted during the
decay of *®T1. The radiation level from this decay could prove to be a debilitating hazard to the patient and
an unacceptable risk to the patient’s family members and the medical staff involved in the treatment. There
are also particular concerns about the long-term dose levels to medical personnel who treat multiple
patients.

A potential solution to this dilemma is the use of *"*Bi produced from the decay chain of ***U
(Pippen 1995). Bismuth-213 has the unique properties of being primarily an alpha emitter (by way of
*13Po) and having only a 2% probability of decaying to **T1, which emits a 1.5-MeV gamma-ray. This
compares with 36% probability for *'*Bi to decay to ***T1, which emits a 2.6-MeV gamma ray. Still, it is

chemically identical to *’Bi, with a similar half-life.

B.2 BISMUTH RECOVERY PROCESS
Recovery of *'*Bi involves a three-step process, as shown in Fig. B.1.

e Recovery of **’Th from *U. The **U is dissolved in acid, and ***Th and its decay product are
separated from the uranium by ion exchange in a shielded process facility with appropriate
safeguards. The resulting thorium-bearing solution contains essentially no fissile uranium; has no
nuclear weapons use; and, therefore, poses no complications in terms of safeguards or nuclear
critically.

e Recovery of **’Ac from ***Th. The **Ac is separated from **Th and the other decay products.
This chemical separation is done in a shielded process facility. No safeguards are required because
there is no fissile material. Because actinium is not a part of the decay chain of ***U, this
separation removes the undesirable product, ***T1, and its precursors. The **Ac is packaged in a
biomedical generator system and is sent to the hospital.

e Recovery of the *’Bi product. At the hospital, the *’Bi is recovered from the ***Ac, converted into
the appropriate chemical form, and injected into the patient. The *'*Bi has a short half-life
(45.6 minutes). The transport times are too long for separation of the *’Bi at a central site;
therefore, the final separations and chemical processing are done at the hospital.

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B-5

Currently, **Th produced from the decay of **U is the only source of ?*Bi. Additionally, **Th could
be produced by irradiation of ***Ra in a nuclear reactor (Feinendegen and McClure 1996). This alternative
production route would allow added production of *’Bi; however, the levels of the contaminant, ***Th,
produced by irradiation of radium, are much higher than those from decay of **U/*?U mixtures in

inventory.

B.3 RECOVERY OF THORIUM-229 FROM URANIUM-233

After 30 years of storage, 1 kg of U will contain ~20 mg of ***Th. The ***U currently stored at
ORNL contains ~40 g of **Th (Fienendegen and McClure 1996). Thorium-229 can be recovered from the
uranium as a separate process or as a step in an isotopic dilution process using the aqueous nitrate blend
method. Recovery of the ***Th involves dissolving the uranium in nitric acid and passing the solution
through an ion-exchange column to remove the thorium. After the recovery step, the **U nitrate solution
could be either converted back to oxide and stored in standardized packages for future use or disposal or
diluted with DU solution prior to conversion to oxide. The entire process may be repeated after several
years for in-growth of *Th and other decay products. It is likely that the isotopic dilution of the **U
would have little affect on this application. The decay chain of ***U, which would be used as the blend-
down material, does not contain actinium. Therefore, the third separation step in the recovery of *"’Bi
would still isolate the desired part of the ***U decay chain.

Separation of *Th from ***U has been successfully demonstrated on a small scale using an anion ion-
exchange resin. The separation is accomplished by first dissolving the ***U in nitric acid to give a
concentration of 200 to 400 g U/L in 8 to 9 M nitric acid solution. The solution is then passed through an
anion ion-exchange resin, during which time the thorium and a portion of the uranium collect on the resin.
The uranium is washed from the resin with a volume of 8 M nitric acid, which is about equal to the volume
of solution processed. The ***Th is then recovered by washing the ion-exchange resin with 0.1 M nitric
acid. Further processing would be required to prepare the **Th for medical use as previously described.
Uranium solution from the thorium removal process is in a form suitable for isotopic dilution by the
aqueous nitrate blending process (ORNL 1995) and/or conversion to oxide for storage or disposal as
previously described. The process will work on ***U-containing neutron absorbers such as cadmium and

gadolinium.
B-6

B.4. REFERENCES FOR APPENDIX B

Bereolos, P. J., L. C. Lewis, C. W. Forsberg, S. N. Storch, and A. M. Krichinsky, June 1998. Draft:
Strategy for Future Use and Disposition of Uranium-233: History, Inventories, Storage Facilities, and
Potential Future Uses, ORNL/TM-13551, Lockheed Martin Energy Research Corp., Oak Ridge National
Laboratory, Oak Ridge, Tennessee.

Fienendegen, L. E. and J. J. McClure (eds.), May 1996. Workshop: Alpha-Emitters for Medical Therapy,
Denver, Colorado, May 30-31, 1996, DOE/NE-0113, U.S. Department of Energy,
Germantown, Maryland.

Knapp, Jr., F. F. and S. Mirzadeh, 1994. “The Continuing Important Role of Radionuclide Generator
Systems for Nuclear Medicine,” Eur. J. Nucl. Med., 21(10), 1151.

Geerlings, M. W., R. van der Hoot, F. M. Kasperson, and C. Apostolides, 1993. “The Feasibility of **Ac
as a Source of Alpha-Particles in Radioimmunotherapy,” Nuclear Medical Communications, 14(2):121.

Oak Ridge National Laboratory, July 13, 1995. Storage and Disposition of Weapons-Usable U-233 by
Aqueous Nitrate Blending with Natural or Depleted Uranium to 2% Enrichment in the Radiochemical
Development Facility at the Oak Ridge National Laboratory, ORNL/MD/LTR-25, Rev. 0, Lockheed
Martin Energy Systems, Oak Ridge National Laboratory, Oak Ridge, Tennessee.

Pippen, C. G, et al., 1995. “Recovery of *’Bi from an ** Ac Cow: Application to the Radiolabeling of
Monoclonal Antibodies with Alpha-Emitter ?°Bi,” Proc. 206th Am. Chem. Soc. Mtg.,
August 22-27, 1993, Chicago, Illinois, American Chemical Society, Washington, D.C.
Appendix C

RECOVERY OF URANIUM FROM CEUSP CANS

C-1
C.1. INTRODUCTION

Approximately one-half the inventory of ***U is stored in CEUSP cans, including the single largest
batch of ***U and another batch of ***U. This storage can is unique among packages used for fissile
material storage. Most fissile materials are stored as loose solids in cans. The uranium in CEUSP cans is
stored as a solid, monolithic U;Oq integral with the can. The can may be cut open, but the uranium will not
fall out. As such, it is an excellent storage form. However, special methods are required for removal from
the can, and most of the disposition options require such removal. Aqueous dissolution removal of ***U

from CEUSP cans is described herein. It is the only demonstrated technology for emptying these cans.

C.2 FILLING OF CEUSP CANS

The CEUSP cans are part of a process developed and used at ORNL to convert liquid solutions of #*U
nitrates into monolithic solids for storage. The uranium nitrate solutions contained uranium, gadolinium,
and cadmium nitrates. The gadolinium and cadmium were used as neutron absorbers in the liquid solutions
to prevent nuclear criticality in storage. In the solidification process, a stainless steel CEUSP can is placed
in a vertical tube furnace at ~800°C. The uranium nitrate solution is slowly added to the can. In the can,
the solution evaporates to dryness, the nitrates decompose to oxides, and a monolithic solid product integral
with the can 1s produced. The uranium product is U;Oq. After solidification, a lid is welded onto the can to
provide a sealed storage can. The can is a relatively robust form of storage, and many of its design
characteristics are dictated by the need for the can to withstand high-temperature processing operations.

Figure C.1 shows a full CEUSP can.

C.3 REMOVAL OF URANIUM FROM CEUSP CANS

A method has been demonstrated by which the U0, CdO, and Gd,O; monoliths contained in the
CEUSP cans can be removed by dissolution with nitric acid in the product can. The removal method was
originally used to (1) dissolve and recycle uranium solutions during development of the CEUSP process
and (2) recycle CEUSP cans during development of the CEUSP process.

The CEUSP material is contained in stainless steel cans (3-in.-diam by 24-in.-high with an internal
volume of about 3 L). Each can contains ~4.1 kg of oxide with an ~50 vol % void fraction. Initially, the

can has about 1.5 L of void space to which nitric acid can be added to dissolve the oxide.

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C-5

In-can dissolution tests were conducted using surrogate CEUSP product made from DU, cadmium, and
gadolinium oxides by the same method as that used in the CEUSP process. The cans were opened at the
top, and an air-sparge tube was inserted into the void space for agitation. The void space was filled with
6 to 8 M nitric acid at room temperature. In these tests, the off-gas from the can, which did not appear to
cause problems, was sent to the cell exhaust system. After reacting for a predetermined time, the acid was
decanted, and fresh acid was added until dissolution was completed. Approximately 1 to 2 d were required

to dissolve the material under the conditions used.
Appendix D

CURRENT TECHNICAL CONSTRAINTS FOR WIPP
D.1 INTRODUCTION

This appendix is a discussion of the technical constraints and limitations on the disposition of ***U-
bearing materials in a geologic repository like WIPP.

For disposal, TRUW containing ***U is acceptable at WIPP if it meets the WAC documented in the
latest version of the WIPP WAC report (DOE Carlsbad Area Office April 1996). For a waste to be
certified as TRUW, its TRU content must exceed 100 nCi/g. If that requirement is met, other radionuclides
such as **U may be present as co-contaminants. A waste stream with only U as a contaminant would
not be acceptable for WIPP. As a provision of DOE Order 5820.2A (DOE September 26, 1988) indicates,
sites may elect to handle (i.e., treat, transport, and store) wastes contaminated with **U as TRUW based
on the potential hazards. However, according to the WIPP Land Withdrawal Act (U.S. Congress 1996),
such materials cannot be shipped to WIPP for disposal.

Also, a significant inventory of wastes exist containing ***U with no TRU radionuclides
(Forsberg, Storch, and Lewis July 7, 1998). Although these wastes cannot go to WIPP, the WIPP WAC
document provides the best current (but not legally binding) definition of waste form and packaging
requirements for eventual disposal of ***U wastes. WIPP is designed for wastes for which the primary
hazard is alpha radiation. The primary hazard of **U is alpha radiation; thus, WIPP is technically suitable
for accepting ***U wastes and has suitable WAC for this purpose.

By law (U.S. Congress 1996), WIPP is authorized to receive and dispose of 175,600 m’ of TRUW
generated from defense operations. In its EIS for WIPP (DOE September 1997), DOE has determined that
another 142,000 m’ of wastes may require disposal by 2033 in a geologic disposal facility such as WIPP.
Much of this waste does not yet exist in packaged form; it will be generated as old cold-war facilities are
decontaminated and decommissioned. These wastes include added defense TRUW, other government
TRUW generated from nondefense activities, and other wastes that may require geological disposal. The
other wastes include existing wastes containing ***U that are in storage—primarily at ORNL and INEEL
(Forsberg, Storch, and Lewis July 7, 1998). Because of the 142,000 m’ of other wastes that ultimately will
require disposal, (1) a second WIPP-type facility must be built, (2) WIPP must be expanded with

congressional authorization to accept these other materials, or (3) some new option must be identified.

D-3
D-4
D.2 WACs FOR 22U

Based on the WIPP WAC, Table D.1 lists the controlling WAC for a ***U waste authorized for
disposal in a WIPP-type repository. Section D.3 provides a more detailed description of the WIPP WAC
that were used to derive Table D.1. Three criteria limit the amount of *°U that can be placed in a 55-gal
drum:

»  Weight (or mass). Total drum weight (i.e., weight of both the drum itself and the waste content) is
limited by the capacity of WIPP waste-handling equipment and vehicle-weight limits.

*  Nuclear criticality. The WIPP facility uses a criticality control strategy dependent upon mass
limits of fissile materials in WPs. For CH waste, this limit is 200 g ?*U in a 55-gal drum. In the
current design of the RH waste canister, which can hold up to three 55-gal drums, there is also a
container limit of 325 g. This limit 1s imposed by the specific transport vehicle and is not a
repository limit. With an alternative transport system, each 55-gal drum of RH waste could accept
200 g of *U.

The WIPP facility is technically designed for alpha wastes, which include *°U. Unlike plutonium
and most other fissile materials, ?**U can be isotopically diluted with DU as an alternative method
to eliminate criticality problems. The WIPP WAC do not consider isotopic dilution as a criticality
control strategy because no significant quantities of “*U were expected to be disposed of in this
facility. However, this alternative criticality control strategy, which is applicable only to fissile
uranium isotopes, eliminates all mass criticality limits for ***U.

» Alpha activity limits. The WIPP WAC limit the quantity of alpha materials per container with
different limits for untreated vs treated wastes. These criteria are designed to limit the potential
consequences of certain types of accidents.

As is evident from Table D.1, WIPP criticality criteria control the maximum quantities of U in a WP
if mass limits are used as a criticality control strategy for *U. This limit is 200 g of Z*U per drum. If the
*¥U is isotopically diluted with DU, this limit no longer applies.

If criticality criteria do not control the amount of **U in a WP, other WAC would limit the ***U per
55-gal drum. For untreated CH waste, the alpha activity limit of 1.3 kg ***U/55-gal drum is limiting. For
CH-treated waste, the maximum container weight limit, ~450 kg/55-gal drum, controls the ***U content per
drum. To control nuclear criticality by isotopic dilution, ~ 188 kg of DU (0.2 wt % ***U) is required per
kilogram of ***U. If the drum weight limit is ~450 kg and the waste form is assumed to be solid uranium
metal, the drum can contain <2.2 kg of ’U. For RH wastes, the weight limits and alpha activity limits are
by coincidence almost identical with the slightly more restrictive alpha activity limits, which restrict the

*¥U per container to 16 kg U or 5.3 kg per equivalent 55-gal drum.
D-5

Table D.1. Summary of WIPP WAC as applied to 23U

 

 

 

Waste category
Constraint CH RH

Container 55-gal drum RH container (accepts three 55-gal drums)
Weight limit (gross) 450 kg 3,600 kg
Criticality control strategy

Mass limits 200 g/drum 325 g/container

Isotopic dilution No limit No limit
Activity limits

Untreated waste 1.3 kg *U 16 kg 2*U

Solidified waste 29 kg #°U 16 kg 2*U
Radiation dose limit <200 mrem/h >200 mrem/h and <1,000 mrem/h

 

D.3 WIPP WAC

The WIPP WAC document identifies the requirements that must be met by participating sites before
any TRUWs are shipped for disposal to the WIPP facility. As stated in the “Executive Summary” of the
WIPP WAC document, “The WIPP Project will comply with applicable federal and state regulations and
requirements, including those in Titles 10, 40, and 49 of the Code of Federal Regulations (CFR). The
WAC document, DOE/WIPP-069, serves as the primary directive for assuring the safe handling,
transportation, and disposal of defense TRU wastes in the WIPP and for certification of these wastes. The
WAC document identifies the requirements that must be met by participating Sites before these TRU
wastes may be shipped for disposal in the WIPP facility.”

As a consequence of this commitment, the WIPP Project is required to comply with the following title
sections of the CFR:

 NRC and DOE standards in CFR Title 10

e EPA standards in CFR Title 40
e DOT standards in CFR Title 49.
D-6

The WIPP WAC reports separate criteria for CH and RH TRUWSs. For RH TRUW, the WACs
reported in the WIPP WAC document are preliminary in nature. For CH TRUW, all weight criteria must
be met, and these are summarized in Table D.2. WIPP WAC packaging configurations, authorized for CH
TRUW shipments, are given in Table D.3. Different payload configurations are restricted by different
weight requirements for CH TRUW payload containers, TRU package transporter Type Il (TRUPACT-II)
shipments, and loaded TRUPACT-IIs. For example, a payload assembly of fourteen 55-gal drums may not
exceed 3,300 kg (7,265 1b) even through the maximum weight of a 55-gal drum may be 450 kg (1,000 Ib).

The WIPP WAC also mentions that although the maximum weight of a payload assembly must not
exceed 3,300 kg (7,265 1b), the weight available for the CH TRUW payload assembly will be less
depending on the as-built weight of a TRUPACT-II to be used. [The average as-built weight of production
TRUPACT-II’s 1s 5,670 kg (12,705 1b).] The weight available for the CH TRUW waste payload assembly
is obtained by subtracting the as-built weight of a TRUPACT-II from the maximum gross weight, which as
Table D.2 shows, 1s 8,730 kg (19,250 1b). The maximum gross weight per TRUPACT-II is based on an
approximate as-built weight of 5,920 kg (13,050 Ib) and an average payload weight of 2,810 kg (6,200 Ib),
which is usually the limiting weight for two TRUPACT-II’s per shipment. DOT’s limit of 36,300 kg
(80,000 1b) gross vehicle-weight rating must also be met; this is the limiting weight for three TRUPACTsSs
per shipment.

Several criteria reported in the WIPP WAC are based on parameters normalized to the radiological
properties of “’Pu. A comparison of some pertinent radiological properties of uranium isotopes with those
of **’Pu is provided in Table D.4. This table indicates that the fissile gram equivalent (FGE) of **U is 0.9
that of ’Pu (Chapman July 23, 1998). However, Appendix 1.3.7 (Table 10.1) of the TRUPACT-II safety
analysis report (SAR) (DOE Carlsbad Area Office 1994) indicates a value of 1.00 for the FGE of both
*¥U and *®Pu. To be conservative, the latter value (expressing >*U’s equivalence with ***Pu) is used for
the FGE of ***U in this assessment.

Several TRUW acceptance criteria specified in the WIPP WAC document directly impact the
thresholds of **U-bearing waste materials. Separate summaries of these requirements are provided for CH
TRUW in Table D.5a and for RH TRUW in Table D.5bh. The criticality data for CH-TRUW in
Table D.5a are based on the WIPP WAC specifications for the maximum allowable fissile material for the
standard shipping container (TRUPACT-II). These specifications are given for various payload containers
in Table D.6.

As previously indicated, the RH TRUW requirements listed in Table D.5b are preliminary in nature.
Current plans are to ship RH TRUWSs in a specially designed container called the RH TRU 72-B Cask.
The design specifications and requirements of this shipping package are reported in the current SAR for the
RH TRU 72-B Shipping Cask (DOE Carlsbad Area Office September 1996).
D-7

Table D.2. WIPP CH TRUW container and assembly weight criteria®

 

Maximum gross weight

 

Component Ib kg

 

Individual payload container

55-gal steel drum (DOT specification 17C) <1,000 <450
55-gal steel drum (DOT specification 17H) <1,000 <450
55-gal steel drum (UN/1A2/X320/S) <700 <320
55-gal steel drum (UN/1A2/X325/S) <716 <325
55-gal steel drum (UN/1A2/X400/S) <882 <400
55-gal steel drum (UN/1A2/X425/S) <937 <425
55-gal steel drum (UN/1A2/X430/S) <948 <430
55-gal steel drum (UN/1A2/X435/S) <959 <435
55-gal steel drum overpacked in a standard waste box <1,450 <660
(SWB)

SWB? <4,000 <1,810
10-drum overpack (TDOP) <6,700 <3,040

Pipe overpack payload container
Pipe overpack 6-in.-diam <328 <150
Pipe overpack 12-in.-diam <547 <250

Payload container assembly

Payload container assembly <7,265 <3,300
TRUPACT-11 <19,250 <8,730
Truck (tractor/trailer) < 80,000 <36,300

 

“Based on Table 3.2.2.2 of the WIPP WAC document (DOE Carlsbad Area Office April 1996).
PSWB = standard-waste box.
D-8

Table D.3. Maximum number of CH TRUW containers per TRUPACT-II and

authorized packaging configurations”

 

 

Number Type of container(s) per TRUPACT-II
14 55-gal drums
14 pipe overpacks in 55-gal drums
2 SWBs
2 SWBs, each containing one bin
2 SWBs, each containing four 55-gal drums

TDOP, containing ten 55-gal drums
TDOP, containing one SWB
TDOP, containing one bin within a SWB

TDOP, containing four 55-gal drums with an SWB

 

“Based on Table 3.2.1.2 of the WIPP WAC document (DOE Carlsbad Area Office April 1996).
Table D.4. Comparison of uranium isotope and ***Pu characteristics

D-9

a

 

 

 

Radionuclide

Characteristic Unit(s) 320 3y 350 2380 2Py
Half-life Years 6.89E+01 1.59E+05 7.04E+08 4.47E+09 2.41E+04
Specific activity Ci/g 2.14E+01 9.68E-03 2.16E-06 3.36E-07 6.22E-02
Specific power Wi/g 6.73E-01 2.77E-04 5.86E-08 8.38E-09 1.88E-03
Py FGE? g 0 1.00 1.00 0 1.00
Subcritical mass limit g 0 500 700 0 450
L(39)/L(i)° 0 0.90 0.64 0 1.00
9Py E-Ci Ci 80 390 430 430 100
Py E-g g 3.7 40,290 1.99E+08 1.28E+09 1,610
Lung class? W, Y Y Y Y Y W
Weighting factor 0.8 3.9 4.3 4.3 1.0

 

“Adapted from Chapman July 23, 1998; Carlsbad Area Office 1994; and DOE Office of Environment, Safety, and Health

July 1988.

Plutonium-239 FGE, as reported in Appendix 1.3.7 of the TRUPACT-II Safety Analysis Report for Packaging
(DOE Carlsbad Area Office 1994). The ***Pu FGE is an isotopic mass of a radionuclide that has been normalized to ***Pu.
“Ratio of the **°Pu subcritical mass limit to the subcritical mass limit of a particular nuclide. Used for calculating **Pu

FGE.

“Expressed as weekly (W) or yearly (Y).
D-10

Table D.5a. Summary of CH TRUW acceptance criteria requirements that pertain to

233U waste thresholds

 

Requirement

As documented in WIPP WAC*

Metric or >*U equivalent

 

Container criteria
Description

Container/assembly weight

Container weight limit

Nuclear criteria

Criticality (***Pu FGE/container)”

Activity (PE-Ci/container)*

Untreated waste

Solidified/vitrified waste

Thermal power

TRU-alpha activity?

Contact dose rate

DOT Type A 55-gal drums or SWBs

<1,000 1b/55-gal drum
<4,000 1b/SWB
<TRUPACT-II weight limits

<60 1b/55-gal drum

<200 g/55-gal drum
<325 g/SWB
<TRUPACT-II limits

<80 PE-Ci/55-gal drum
<130 PE-Ci/SWB
<1,800 PE-Ci/55-gal drum overpacked
in
SWB or TDOP

<1,800 PE-Ci/55-gal drum

Report if >0.1 W/ft?
<40 W per TRUPACT-II

>100 nCi/g (waste matrix)

<200 mrem/h

DOT Type A 55-gal drums or SWBs
<450 kg/55-gal drum

<1,800 kg/SWB

<TRUPACT-II weight limits

<27 kg/55-gal drum

<200 g/55-gal drums
<325 g/SWB
<TRUPACT-II limits

<1.3 PE-kg/55-gal drum

<2.1 PE-kg/SWB

<29 PE-kg/55-gal drum overpacked in
SWB or TDOP

<29 PE-kg/55-gal drum

Report if >3.5 W/m’
<145 kg 23U per TRUPACT-II

>100 nCi/g (waste matrix)

<200 mrem/h

 

“As specified in the WIPP WAC document (DOE Carlsbad Area Office April 1996).

bSee also Table E.6.

“Radioactivity (per container) expressed in units of ***Pu equivalent curies.
?Activity of alpha-emitting TRU isotopes with half-lives >20 years.
D-11

Table D.5b. Summary of RH TRUW preliminary acceptance criteria requirements that

pertain to **U waste thresholds

 

As documented in

 

TRU-alpha activity?

Contact dose rate

Per canister

Per cask

>100 nCi/g (waste matrix) and
<23 Ci/L

<1,000 rem/h
Preapproval for >100 rem/h

<200 mrem/h

Requirement WIPP WAC* or SAR? Metric or ***U equivalent
Container criteria
Description DOT Type A RH canister® DOT Type A RH canister®
Weight limit for canister and <&,000 Ib <3,630 kg
contents
Weight of empty canister 1,762 1b 800 kg
Weight limit for waste contents <6,238 1b <2,830 kg
in canister
Nuclear criteria
Criticality (*Pu FGE/cask) <325¢ <325¢
Activity (***Pu-E/canister) <1,000 PE-Ci 16 PE-kg
Thermal power (canister) <300 W <1,080 kg **U

>100 nCi/g (waste matrix) and
<2.3 kg 3U/L

<1,000 rem/h
Preapproval for >100 rem/h

<200 mrem/h

 

“As specified in the WIPP WAC document (DOE Carlsbad Area Office April 1996).

bAs specified in the SAR for the RH-TRU 72-B shipping cask (DOE Carlsbad Area Office September 1996).
A shipping container that holds three 55-gal drums.

?Activity of alpha-emitting TRU isotopes with half-lives >20 years.
D-12

Table D.6. Maximum allowable fissile material, expressed as 2°Pu FGE
for CH TRUW in the TRUPACT-II*

 

 

Payload container Mass of maximum allowable fissile material (**°Pu FGE)
55-gal drum 200
SWB 325
TDOP 325
Pipe component overpacked in a 55-gal drum 200
TRUPACT-II 325
TRUPACT-II (14 pipe overpacks) 2,800

 

“Adapted from Table 3.3.1.2 of the WIPP WAC document (DOE Carlsbad Area Office April 1996).

A comparison of WIPP-based nuclear criteria for selected radionuclides is provided for CH TRUW
and RH TRUW in Tables D.7a and D.7b, respectively. Each table lists separate criteria for Z*Pu and four
uranium isotopes: **U, *U, #*°U, and ***U.

A sample sensitivity analysis is provided in the following to show the impacts of the WIPP WAC and
isotopic dilution on ***U packaging in 55-gal drums. Four cases (A, B, C, and D) are considered, one of
which shows the amount of ***U (in the form of 2*UQ,) that could be placed into a 55-gal drum if the **U
content were isotopically diluted to 0.53 wt % in DU, which has 0.20 wt % **U. This assessment is made
to show also the quantity of **U (in the form of *?UQ,) that could be placed into a 55-gal drum under four
different conditions. For the four cases considered, the results are summarized in the following for a

nominal 1 t domestic ***U inventory:
D.3.1 Case A
WIPP WAC for CH TRUW as **UQ, in a 55-gal drum:
* Controlling factor
—  Nuclear criticality limit: 0.200 kg (as *°U) = 0.227 kg (as **UO,).

e Resulting number of 55-gal drums needed from a 1,000-kg ***U inventory = 1,000 kg/0.200 kg
*¥U per drum = 5,000 drums.
D-13

Table D.7a. Comparison of WIPP-based nuclear criteria for various radionuclides

 

 

 

comprising CH TRUW*
Radionuclide
Nuclear criteria Units il U] 2y 25U 28y 2¥Pu
Maximum criticality limits
55-gal drum *Pu FGE (g) 0 200 200 0 200
SWB *Pu FGE (g) 0 325 325 0 325
TRUPACT-II *Pu FGE (g) 0 325 325 0 325
TRUPACT-II (with 14 pipe overpacks) 0 2,800 2,800 0 2,800
Maximum activity limits
Untreated waste
55-gal drum PE-Ci 64 312 340 340 80
SWB PE-Ci 100 510 560 560 130
TDOP PE-Ci 1,440 7,020 7,740 7,740 1,800
Solidified/vitrified waste
55-gal drum PE-Ci 1,440 7,020 7,740 7,740 1,800

 

“Adapted from the WIPP WAC document (DOE Carlsbad Area Office April 1996).
D-14

Table D.7b. Comparison of WIPP-based nuclear criteria for various radionuclides

 

 

 

comprising RH TRUW*
Radionuclide
Nuclear criteria Units il U] 2y 25U 28y 2¥Pu

Maximum criticality limits

55-gal drum »%Pu FGE (g) 0 200 200 0 200

SWB »%Pu FGE (g) 0 325 325 0 325

RH TRU 72-B shipping cask”  **Pu FGE (g) 0 325 325 0 325
Maximum activity limits

Untreated waste PE-Ci NA® NA NA NA NA

Solidified/vitrified waste PE-Ci NA NA NA NA NA

 

“Adapted from the WIPP WAC document (DOE Carlsbad Area Office April 1996).
’Based on SAR for RH TRU 72-B shipping cask (DOE Carlsbad Area Office September 1996).
‘NA = not applicable (i.e., no activity limits are specifically indicated for untreated and treated wastes).

D.3.2 Case B

Isotopic dilution of solidified CH TRUW as ***UQ, in a 55-gal drum (if the ***U is considered untreated

waste, the mass limit per drum of U is 450 kg):
* Controlling factor

— Maximum permissible mass of drum contents (after subtracting drum weight): 426.4 kg (as
UO,) = 375.8 kg (as U).

e Mass of **U per isotopically diluted drum = 0.0053 x 375.8 kg = 1.992 kg.

e  Resulting number of 55-gal drums needed from a 1,000 kg ***U inventory = 1,000 kg/1.992 kg ***U
per drum = 502 drums.
D.3.3 Case C

WIPP WAC for RH TRUW as *’UQ, in two or three 55-gal drums in an RH TRU 72-B shipping

canister:
* Controlling factor
—  Nuclear criticality limit: 0.325 kg (as *°U) = 0.369 kg (as *’UQ,).

« Restriction: a single drum can contain <0.200 kg **U. The remaining mass of 0.125 kg *°U
allotted for an RH TRU 72-B shipping canister must go into one or two more drums.

* Assuming only one more drum is needed, then the resulting number of 55-gal drums needed from a
1,000 kg ***U inventory = 1,000 kg/0.325 kg ***U per 2 drums = 6,154 drums.

D.3.4 CaseD
Isotopic dilution of RH TRUW as ***UQ, in three 55-gal drums in an RH TRU 72-B shipping canister:
* Controlling factor

— Maximum permissible mass of an RH canister containing three 55-gal drums (gross canister
weight limit is 3630 kg; canister weight is 800 kg): 2830 kg (as UO,) =2494.5 kg (as U) =
831.5 kg (as U per 55-gal drum).

e Mass of **U in an isotopically diluted drum = 0.0053 x 831.5 kg = 4.407 kg.

« Resulting number of 55-gal drums needed from a 1,000 kg ***U inventory:

= 1,000 kg/4.407 kg/drum = 227 drums.

For both CH and RH wastes, these results show that isotopic dilution results in very significant
reductions in the number of 55-gal drums that would be required for the disposition of a domestic *°U
inventory. In the previous examples, isotopic dilution results in a requirement of nearly 4,500 fewer drums
for CH wastes (Case B vs Case A) and nearly 5,930 fewer drums for RH wastes (Case D vs Case C). The
cost savings associated with these potential differences can be shown to be very significant.

The WIPP Disposal Phase Final EIS (DOE Carlsbad Area Office September 1997) indicates the
following costs [in millions (M) of 1994 dollars ($)]:

» Total life-cycle cost for TRUW program: $19,030M.
» Total life-cycle costs for waste treatment: $12,140M.
D-16

The difference between Items 1 and 2 above gives an estimated total cost of $6,890M for waste
transport and disposal. This cost is for about 168,500 m’ of waste. Expressing this cost on a unit volume
basis gives ($6,890M)/(168,500 m®) = $40,890/m’. Since the volume capacity of a 55-gal drum is
0.2082 m?’, this unit cost can be expressed as ~$8,500/drum. Applying this result to the drum inventory
savings previously indicated gives a total potential cost savings of ~$38M for a domestic ***U inventory of

CH wastes and about $50M for a domestic ***U inventory of RH wastes.

D.4 GLOSSARY OF TERMS FOR APPENDIX D
Canister gross weight: Total weight of a waste canister and its contents.

Dunnage: Loose packing material used to protect a waste container’s contents from damage during
transport.

FGE (**Pu): An isotopic mass of a radionuclide that has been normalized to **°Pu.

Overpack: A payload container placed around another container to control contamination, or enclose a
damaged container.

Package: The reusable Type B shipping container (i.e., TRUPACT-II or RH-TRUW 72-B Cask) loaded
with TRUW payload containers, which has been prepared for shipment in accordance with the TRUW
Packaging QA Program.

Packaging: The reusable Type B shipping container for transport of TRUW payload containers

(i.e., TRUPACT-II or RH-TRUW 72-B Cask). A transportation device consisting of an assembly of
components necessary to ensure compliance with the requirements of Titles 49 CFR Part 173, Subpart 1
(“Shippers-General Requirements and Packaging”) and 10 CFR Part 71 (“Packaging and Transportation of
Radioactive Material”).

Payload container: The outermost container for TRUW material that is placed in a reusable Type B
shipping container (i.e., TRUPACT-II or RH-TRUW 72-B Cask) for transport.

Payload container assembly: An assembly of payload containers, such as a seven-pack of drums, which
1s intended to be handled and emplaced as a single unit.

Plutonium equivalent curie (PE-Ci): An equivalent radiotoxic hazard (radioactivity) of a radionuclide
normalized to **Pu.

Standard waste box (SWB): A payload container authorized for use with TRUPACT-II Transportation
Packages that meet DOT Specification 7A Type A.

Ten drum overpack (TDOP): A specialized payload container authorized for use within the
TRUPACT-II packaging that meets DOT specification 7A Type A.
D-17

Thermal power (or decay heat): A measure of the rate of heat-energy emission that results from the
radioactive decay of a material. A unit of thermal power commonly used is the watt (W).

Transuranic waste (TRUW): Radioactive waste that, at the time of assay, contains >100 nCi/g of alpha-
emitting isotopes with atomic numbers >92 and half-lives >20 years.

TRUW, contact-handled (CH): Packaged TRUW with an external contact dose rate at the container
surface that is <200 mrem/h.

TRUW, remote-handled (RH): Packaged TRUW with an external contact dose rate at the container
surface that is >200 mrem/h. For WIPP, there is an upper limit of 1000 rem/h.

TRU alpha activity: Radioactivity (Ci) from the alpha radiation of TRU radionuclides.

TRUPACT-II: Transuranic Package Transporter II; An NRC-certified, Type B transportation packaging
used for the shipment of CH TRUW.

D.5 REFERENCES FOR APPENDIX D

Chapman, J. A., July 23, 1998. Oak Ridge National Laboratory, Oak Ridge, Tennessee, “WIPP
Parameters,” correspondence to Stephen N. Storch, Oak Ridge National Laboratory,
Oak Ridge, Tennessee.

Forsberg, C. W., S. N. Storch, and L. C. Lewis, July 7, 1998. Uranium-233 Waste Definition: Disposal
Options, Safeguards, Criticality Control, and Arms Control, ORNL/TM-13591, Lockheed Martin Energy
Research Corp., Oak Ridge National Laboratory, Oak Ridge, Tennessee.

U. S. Congress, 1996. Waste Isolation Pilot Plant Land Withdrawal Act, Public Law 102-579, as
amended.

U. S. Department of Energy, Carlsbad Area Office, 1994. Safety Analysis Report for the TRUPACT-II
Shipping Package (SARP), U.S. NRC Dockett No. 71-9218, DOE/WIPP-97003692,
Carlsbad, New Mexico.

U. S. Department of Energy, Carlsbad Area Office, April 1996. Waste Acceptance Criteria for the Waste
Isolation Pilot Plant, DOE/WIPP-069, Rev. 5, Carlsbad, New Mexico.

U. S. Department of Energy, Carlsbad Area Office, September 1996. Safety Analysis Report for the
RH-TRU 72-B Shipping Cask, U.S. NRC Dockett No. 71-9212, Rev. 0, Carlsbad, New Mexico.

U. S. Department of Energy, Carlsbad Area Office, September 1997. Waste Isolation Pilot Plant
Disposal Phase Final Supplemental Environmental Impact Statement, DOE/EIS-0026-S-2,
Carlsbad, New Mexico.

U. S. Department of Energy, Office of Environment, Safety, and Health, July 1988. Internal Dose
Conversion Factors for Calculation of Dose to the Public, DOE/EH-0071, Washington, D.C.
D-18

U. S. Department of Energy, September 26, 1988. DOE Order 5820.2A, Radioactive Waste Management,
Washington, D.C.
14-15.
16.
17.
18-22.
23.
24.
25.
26-30.
31.
32-34.
35.
36.

64.

65.

66.

67.

68.

69.

ORNL/TM-13553

INTERNAL DISTRIBUTION

C. W. Alexander 37. G.T.Mays

E. C. Beahm 38. L. E. McNeese

J. M. Begovich 39-41. G. E. Michaels

P. J. Bereolos 42. H.J. Monroe

L. F. Blankner 43. D. L. Moses

H. E. Clark 44. R.J. Moses

E. D. Collins 45. B. D. Patton

S. O. Cox 46. D. W. Ramey

A. G. Croff 47. D. E. Reichle

L. R. Dole 48. K. D. Rowley

W. D. Duerksen 49. J. E. Rushton

K. R. Elam 50. A.R. Sadlowe

C. W. Forsberg 51. B. W. Starnes

E. C. Fox 52-56. S. N. Storch

S. R. Greene 57. J.R. Trabalka

M. J. Haire 58. D. Turner

R. F. Holdaway 59. O.F. Webb

C. M. Hopper 60. K. A. Williams

A. S. Icenhour 61. Central Research Library

A. M. Krichinsky 62. Laboratory Records—RC

S. B. Ludwig 63. Document Reference Section
EXTERNAL DISTRIBUTION

Joe Arango, U.S. Department of Energy, S-3.1, Rm. 6H-025,
1000 Independence Ave., S.W., Washington, D.C. 20585.

M. W. Barlow, Westinghouse Savannah River Co., Savannah River Site,
Bldg. 704-C, Aiken, South Carolina 29808.

W. E. Bickford, Westinghouse Savannah River Co., Savannah River Site,
Bldg. 773-41A, Aiken, South Carolina 29808.

Scott Boeke, Westinghouse Savannah River Co., Savannah River Site,
Aiken, South Carolina 29808.

Mark C. Bronson, Lawrence Livermore National Laboratory, P.O. Box 808,
Livermore, California 94550.

Bill Brummend, Lawrence Livermore National Laboratory, P.O. Box 808,
Livermore, California 94550.
70.

1.

72.

73.

74.

75.

76.

TT.

78.

79.

80.

81.

82.

83.

84.

Mathew Bunn, National Academy of Sciences, Founder’s
Bldg., 1055 Thomas Jefferson Street, N.W., Washington, D.C. 20007.

H. R. Canter, U.S. Department of Energy, Office of Fissile Materials
Disposition, 1000 Independence Ave., S.W., Washington, D.C. 20585.

Alice Caponiti, U.S. Department of Energy, Office of Fissile Materials
Disposition, 1000 Independence Ave., S.W., Washington, D.C. 20585.

Nate Chipman, Idaho National Engineering and Environmental Laboratory,
WCB, MS-3114, P.O. Box 1625, Idaho Falls, Idaho 83415-3114.

George Christian, Lockheed Martin Idaho Technologies Company,
North Building, Suite 1404, 955 L’Enfant Plaza, S.W.,
Washington, D.C. 20024.

John Clouet, TRW Environmental Safety Systems, Inc., M/S 423, 1261 Town
Center Dr., Las Vegas, Nevada 89134.

Ray Cooperstein, U.S. Department of Energy, Mail Stop DP-45,
19901 Germantown Rd., Germantown, Maryland 20874.

A. L. Cygelman, U.S. Department of Energy, Office of Fissile Materials
Disposition, DOE/MD-3, Forrestal Bldg., 1000 Independence Ave., S.W.,
Washington, D.C. 20585.

P. D’Entromont, Westinghouse Savannah River Co., Savannah River Site,
Aiken, South Carolina 29808.

Paul Daniel, U.S. Department of Energy, Yucca Mountain Site
Characterization Office, M/S 523, Suite A, 1551 Hillshire Dr.,
Las Vegas, Nevada 89134.

Bill Danker, U.S. Department of Energy, DOE/MD-3, Forrestal Bldg.,
1000 Independence Ave., S.W., Washington, D.C. 20585.

Wesley Davis, TRW Environmental Safety Systems, Inc., M/S 423,
1261 Town Center Dr., Las Vegas, Nevada 89134.

John Dickenson, Westinghouse Savannah River Co., Savannah River Site,
Aiken, South Carolina 29808.

A. Disabatino, Lawrence Livermore National Laboratory, P.O. Box 808,
Livermore, California 94550.

James Duguid, TRW Environmental Safety Systems, Inc., Suite 800,
2650 Park Tower Dr., Vienna, Virginia 22180.
85.

86.

87.

88.

89.

90.

91.

92.

93.

94.

95.

96.

97.

98.

99.

100.

Bart Ebbinghaus, Lawrence Livermore National Laboratory, P.O. Box 808,
Livermore, California 94550.

Randy Erickson, Los Alamos National Laboratory, MS-F660, P.O. Box 1663,
Los Alamos, New Mexico 87545.

John Evans, U.S. Department of Energy, 1000 Independence Ave., S.W.,
Washington, D.C. 20585.

Roland Felt, 780 DOE Place, Idaho Falls, Idaho 83415-1216.

Dick Frushour, Westinghouse Savannah River Co., Savannah River Site,
Aiken, South Carolina 29808.

Michael L. Gates, U.S. Department of Energy, Project Planning and
Integration, Nuclear Materials Stewardship Project Office, Albuquerque
Operations Office, P.O. Box 5400, Albuquerque, New Mexico 87185-5400.

Arnold Gnevara, U.S. Department of Energy, DOE/EM-66,
1000 Independence Ave., S.W., Washington, D.C. 20585.

Peter Gottlieb, TRW Environmental Safety Systems, Inc., M/S 423,
1261 Town Center Dr., Las Vegas, Nevada 89134.

Tom Gould, Lawrence Livermore National Laboratory, P.O. Box 808,
Livermore, California 94550.

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

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

Reginald Hall, Advanced Integrated Management Service, Suite 203B,
702 S. Illinois Ave., Oak Ridge, Tennessee 37831.

R. Harmon, Lockheed Martin Idaho Technologies, Inc., P.O. Box 1625,
Idaho Falls, Idaho 83415.

Dave Haught, U.S. Department of Energy, Yucca Mountain Site
Characterization Office, M/S 523, Suite A, 1551 Hillshire Dr.,
Las Vegas, Nevada 89134.

Roger Henry, P. O. Box 1625, Idaho Falls, Idaho 83415-3805.
Laura S. H. Holgate, U.S. Department of Energy, Office of Fissile Materials

Disposition, DOE/MD-1, 1000 Independence Ave., S. W.,
Washington, D.C. 20585.
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Brent Ives, Lawrence Livermore National Laboratory, 7000 East Ave.,
Livermore, California 94550.

Krishna Iyengar, TRW Environmental Safety Systems, Inc., M/S 423,
1261 Town Center Dr., Las Vegas, Nevada 89134.

N. C. Iyer, Westinghouse Savannah River Co., Savannah River Site,
Bldg. 773-A, Aiken, South Carolina 29808.

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

Gregory V. Johnson, Westinghouse Savannah River Co., Savannah River Site,
Aiken, South Carolina 29808.

Hoyt Johnson, U.S. Department of Energy, EM-66, Forrestal Bldg.,
1000 Independence Ave., S.W., Washington, D.C. 20585.

Ed Jones, Lawrence Livermore National Laboratory, 7000 East Ave., L-634,
Livermore, California 94550.

J. E. Jones, Jr., Haselwood Enterprises, Inc., Suite 300A,
1009 Commerce Park, Oak Ridge, Tennessee 37830.

S. L. Krahn, Defense Nuclear Facilities Safety Board, Suite 700,
625 Indiana Ave., N.W., Washington, D.C. 20004.

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

Christi Leigh, Sandia National Laboratories, M/S 0778, P.O. Box 5800,
Albuquerque, New Mexico 87185.

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

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

Betty Ahnde Lin, Princeton University, Center for Energy and Environmental
Studies, Engineering Quadrangle, Princeton, New Jersey 08544.

Herbert Massie, U.S. Defense Nuclear Facilities Safety 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.
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Don McWhorter, Westinghouse Savannah River Company, Savannah River
Site, Bldg. 704-F, Aiken, South Carolina 29808.

Ed Moore, Westinghouse Savannah River Company, Savannah River Site,
Bldg. 773-41A, Rm. 125, P.O. Box 616, Aiken, South Carolina 29808.

Ram B. Murthy, U.S. Department of Energy, Yucca Mountain Site
Characterization Office, M/S 523, Suite A, 1551 Hillshire Dr.,
Las Vegas, Nevada 89134.

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

Dave Neiswander, Advanced Integrated Management Services, Inc.,
Suite 203B, 702 S. Illinois Ave., Oak Ridge, Tennessee 37830.

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

Donald Nitti, TRW Environmental Safety Systems, Inc., M/S 423, 1261 Town
Center Dr., Las Vegas, Nevada 89134.

David Nulton, U.S. Department of Energy, Office of Fissile Materials
Disposition, DOE/MD, 1000 Independence Ave., S.W.,
Washington, D.C. 20585.

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

Bill Pearson, Westinghouse Savannah River Co., Savannah River Site,
Aiken, South Carolina 29808.

K. L. Pilcher, Haselwood Enterprise, Inc., Suite 300A, 1009 Commerce Park,
Oak Ridge, Tennessee 37830.

Robert P. Rechard, Sandia National Laboratories, New Mexico,
P.O. Box 5800, Albuquerque, New Mexico 87185-0779

Victor Reis, U.S. Department of Energy, Assistant Secretary, Defense
Programs, 1000 Independence Ave., S.W., Washington, D.C. 20585.

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

Gary D. Roberson, U.S. Department of Energy, NMSPO, P.O. Box 5400,
Albuquerque, New Mexico 87185-5400.

Greg Rudy, U.S. Department of Energy, Bldg. 703-A/E245N, P.O. Box A,
Aiken, South Carolina 29802.
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S. S. Sareen, TRW, Suite 800, 2650 Park Tower Dr., Vienna, Virginia 22180.

Linda Seward, Idaho National Engineering and Environmental Laboratory,
P.O. Box 1625, Idaho Falls, Idaho 83415.

Henry Shaw, Lawrence Livermore National Laboratory, P.O. Box 808,
Livermore, California 94550.

David Stahl, TRW Environmental Safety Systems, Inc., M/S 423, 1261 Town
Center Dr., Las Vegas, Nevada 89134.

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

B. Stevenson, U.S. Department of Energy, Office of Fissile Materials
Disposition, 1000 Independence Ave., S.W., Washington, D.C. 20585.

John Thompson, U.S. Department of Energy, Office of Fissile Materials
Disposition (DOE/MD), 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.

Rich Van Konynenburg, Lawrence Livermore National Laboratory,
P.O. Box 808, Livermore, California 94550.

Don Vieth, 1154 Cheltenham Place, Maineville, Ohio 45039.

C. Vitale-Smith, Lockheed Martin Idaho Technologies, Inc., P.O. Box 1625,
Idaho Falls, Idaho 83415.

Mike Walrath, Idaho National Engineering and Environmental Laboratory,
P.O. Box 1625, Idaho Falls, Idaho 83415.

K. E. Waltzer, Westinghouse Savannah River Co., Savannah River Site,
Bldg. 703-F, Aiken, South Carolina 29808.

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

Wendell L. Williams, U.S. Department of Energy, MD-3, Forrestal Bldg.,
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.

Loong Yong, Advanced Integrated Management Service, Suite 203B,
702 S. Illinois Ave., Oak Ridge, Tennessee 37831.
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162-163.

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

Neil R. Zack, Los Alamos National Laboratory, MS-E541, P.O. Box 1663,
Los Alamos, New Mexico 87545.

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

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