ORNL-6952.txt

ORNL-6952

Uses For Uranium-233: What Should Be Kept for Future Needs?

C. W. Forsberg

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

September 24, 1999

 

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

LIST OF FIGURES . . ..., Vil
LIST OF TABLES . . . ., Vil
ACRONYMS AND ABBREVIATIONS ... ... . 1X
ACKNOWLEDGMENTS . . . . ., X1
PREF ACE Xiii
EXECUTIVE SUMMARY . . .., XV
ABSTRACT .. XX111
I. INTRODUCTION . ... . e 1
1.1 OBIJECTIVES . .. 1

1.2 CHARACTERISTICS OF *°U USE OR DISPOSE DECISIONS ............ 1

1.3  CAVEATS . 1

1.4 ORGANIZATION OF THIS REPORT . ... ... ... ... ... ... ... ... .. ..... 2

2. CHARACTERISTICS AND INVENTORY OF **U ... ... ... .. .. 3
2.1 CHARACTERISTICS . .. . . 3
2.1.1 Radiological ...... ... ... . . .. . ... 4

2.1.2  Nuclear Criticality . ........ ... ... . ... . . .. . . . ... . ... 7

2.1.3 Safeguards ......... . . ... 7

22 INVENTORY ... 8
2.2.1 High-Isotopic-Quality ***U with Limited Chemical Impurities ... ... .. 11

222 High-Isotopic-Quality ***U with Chemical Diluents . ............... 11

2.2.3 Intermediate-Isotopic-Quality U . ... ... ... .. .. ... ... .. .. .. 11

2.2.4 Low-Isotopic-Quality 2*U ... ... ... ... .. . . .. 12

3. PRODUCTION COSTS FOR 23U .. .. e 13
3.1 HISTORIC PRODUCTION COSTS .. ... ... . . . . . ... 13

3.2 CURRENT SALES PRICE AND PRODUCTION COSTS ................. 13

3.3 FUTURE PRODUCTION COSTS ... ... . .. .. 14

111
CONTENTS (continued)

4. USESOF 22U . ...

4.1

4.2
4.3
4.4
4.5

MEDICAL APPLICATIONS ... .

A.T1.1  USe o
4.1.2 Production Methods for 2*Bi . . ... ... ... ... ...

4121 Production from 22U ... ... ... .. . . ...
41.2.1.1 Production Method ....................... ..
4.12.1.2 Thortum-229 Inventory ......................
41.2.1.3 ProductionlIssues ............ ... . ... . .......
41214 Current Status . ............. .. . . . ... ... ...

4.1.2.2 Production from Radium-226 (**Ra) ... ............... ...
41221 Production Methods ....................... ..
4.12.2.2 Radium-226 Availability . ....................
41223 ProductionlIssues ............ ... .. ... . ... ...

4123 Productionfrom?®Ra ........ ... .. . . ... ... ...

4.1.3 Availability of 2PBi ... ...

4.1.3.1 Demand . ...... ... . ...,
4.1.32 ECONOMICS . .. ..\t

414 Assessment and Conclusions . ............. .. . . ... ... ... ... ..

REACTORS FOR DEEP-SPACE AND OTHER SPECIAL MISSIONS . ... . ..
ANALYTICAL MEASUREMENTS . ... ... ... . .
NUCLEAR WEAPONS RESEARCH ...... ... ... ... ... .. ... ... ... ....
REACTOR FUEL CYCLERESEARCH . ....... ... ... . ... ... ... ... ....

451 Application . ... ... ...
4.5.2 Nonproliferation Fuel Cycles .. ........... ... ... ... ... ......

4521 Isotopic Dilution ......... ... ... ... ... .. ... ... ... ...
45272 Radiation .......... ... ... ...
4.5.2.3 Off-Specification Plutonium . ..........................

45231 Quantity ......... ... .. ...
45232 Quality ...... ... ..

4524, Once-Through Fuel Cycles ............................

v

15

15

15
19

19
19
21
21
23
23
23
25
25
25
26

26
27

27

28
31
31
33

33
34

34
35
37

37
37

37
CONTENTS (continued)

4.5.3 Current Nuclear-Power Thorium-Fuel-Cycle Developments . ... ... ... 38
4.5.3.1 Countries with Active Programs . .................... ... 38
4.5.3.2 Once-Through Thorium Fuel Cycles . .................. .. 38
4.532.1 Nonproliferation............................ 38
45322 Uranium Resources ......................... 39
45323 FuelPerformance ........................... 40
4533 U.S. ResearchPrograms .............................. 40
4533.1 RadkowskyReactor ......................... 40
4.5.3.3.2 National Laboratory, University, Fuel Fabricator
Consortium . . .............................. 41
45333 ANL and Purdue University .................. 42
45334 BNL and Purdue University . . ................. 42
454  Accelerator and Fusion Reactor Development . ... ............ .. ... 42
4541 Accelerators . ....... ... ... ... 43
4542 Fusion-FissionHybrids ............................... 43
455 Material Requirements .. ........ ... . ... . .. .. ... ... ... ... ..., 43
456 ASSESSMENT . ... ... ... 45
4.6 OTHER APPLICATIONS . ... ... .. . . . . 45
5. CONCLUSIONS . .. 47
6. REFERENCES ... . . . . 49
Appendix A RADIATION LEVELS FROM U .. ... ... . ... .. ... ... ... ..., A-1
LIST OF FIGURES

Fig. 2.1 Different fissile materials require varying handling procedures .. ....... ... ..
Fig. 22 Gamma exposure for 1 kg of 2°U with 100 ppm of 22U . ... ...... ... ... ... ..
Fig. 4.1  Potential uses for U . .. ... ... .. .. . ..
Fig. 42  2Udecay chain .. .......... ... . . . . .
Fig. 43  Flowsheet for *"’Bi production from ***U for treatment of cancer ... ... . ... ...

Fig. 44  Minimum mass space nuclear power sources for different power levels and
mission duration . . ...

Fig. 4.5  Maximum radiation levels of fresh fuel with 2°U ... .. ... ... ... ... ... .. ..

LIST OF TABLES

Table ES.1 Quality of major batches of *°U ininventory ...........................
Table ES.2 Uranium-233 uses and applicable U categories . .......................
Table 2.1  Characteristics of weapons-usable materials ............................
Table 2.2  Quality of major batches of *°U ininventory ...........................
Table 2.3  Quality of major batches of U in SNF .. ... .. ... ... ... ..........
Table 4.1  Summary: medical applications . ......... ... ... ... ... ... ... ........
Table 4.2  Uranium sources of **Th for medical applications . ......................
Table 4.3  Summary: reactors for deep-space and other special missions . .............
Table 4.4  Summary: analytical measurements .. ............ ... ... ... ............
Table 4.5  Summary: nuclear weapons . ................... . .. ... ... ...
Table 4.6 ~ Summary: reactor fuelcycle . ...... ... .. ... ... . .. ... ... .. ... ...,

Table 5.1  Uranium-233 uses and applicable U categories . .......................

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

ANL
ATR
BNL
Cd

CEUSP
CH
DOE
DU
EURATOM
h
HEPA
HEU
HFIR
HTGR
[AEA
ITU
INEEL

LEU
LWBR
LWR
MeV
mg
MIT
MSRE
mrem

ORNL

Argonne National Laboratory
Advanced Test Reactor

Brookhaven National Laboratory
Cadmium

day

Consolidated Edison Uranium Solidification Program
contact-handled

U.S. Department of Energy

depleted uranium

European Atomic Energy Community
hour

high-efficiency particulate air

high enriched uranium

High-Flux Irradiation Reactor
high-temperature gas cooled reactor
International Atomic Energy Agency
Institut fur Transurane

Idaho National Engineering and Environmental Laboratory
kilogram

liter

low-enriched uranium

light-water breeder reactor

light-water reactor

million-electron volts

milligram

Massachusetts Institute of Technology
Molten Salt Reactor Experiment
millirem

Oak Ridge National Laboratory

1X
ACRONYMS, ABBREVIATIONS, SCIENTIFIC NOTATION, AND UNITS OF MEASURE
(continued)

ppm part(s) per million
R&D research and development
RH remote-handled

S&S safeguards and security
SNF spent nuclear fuel

SRS Savannah River Site

t metric ton

WGP weapons-grade plutonium
Y-12 Y-12 Plant (Oak Ridge)
AC Actinium-225

238py Plutonium-238

%Py Plutonium-239

#1py Plutonium-241

*Ra Radium-225

22°Ra Radium-226

22Ra Radon-220

2057 Thallium-208

Th Thorium-229

2Th Thorium-232

227 Uranium-232

23U Uranium-233

23U Uranium-235

28y Uranium-238

2B Bismuth-213
ACKNOWLEDGMENTS

We would like to thank the following individuals for providing information and review comments.

 

Individual

Affiliation

 

J. Arango

D. E. Beller

P. Bereolos

R. Cooperstein
J. W. Davidson
L. R. Dole

M. J. Driscoll
E. Greenspan

J. S. Herring

M. S. Kazimi
L. Koch

E. Lahoda

L. M. Lidsky
S. McDeavitt
H. Massie

S. Mirzadah
G. P. Smith, Jr.
W. Spetz

M. Todosow
J. Tseng

L. F. P. Van Swan

H. Vantine

U.S. Department of Energy (DOE)

Los Alamos National Laboratory (LANL)
Advanced Integrated Management Services, Inc.
DOE

LANL

Oak Ridge National Laboratory (ORNL)
Massachusetts Institute of Technology (MIT)
University of California, Berkeley

Idaho National Engineering and Environmental
Laboratory

MIT

Transuranium Institute (Germany)
Westinghouse Electric Corporation
MIT

Argonne National Laboratory

U.S. Defense Nuclear Facilities Safety Board
ORNL

ABB Combustion Engineering, Inc.
Framatome Technologies, Inc.
Brookhaven National Laboratory
DOE

Siemens Power Corporation

Lawrence Livermore National Laboratory

 

X1
PREFACE

This report 1s one of several reports which map 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 2?U. It includes required
activities, identifies what major programmatic decisions will be required, and describes the
potential disposition options.

e 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. It 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 2°U to prevent the potential for nuclear criticality under all expected process
and disposal facility conditions.

«  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 **U-containing material is
waste and what 2*’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 options for **U disposition.

Xiii
EXECUTIVE SUMMARY

Since the end of the cold war, the United States has been evaluating what fissile materials to keep for
potential uses and what fissile materials to declare excess. There are three major fissile materials: high-
enriched uranium (HEU), plutonium, and uranium-233 (**°U). Both HEU and plutonium were produced
in large quantities for use in nuclear weapons and for reactor fuel. Uranium-233 was investigated for use
in nuclear weapons and as a reactor fuel; however, it was never deployed in nuclear weapons or used
commercially as a nuclear fuel. Uranium-233 has limited current uses, but it could have several future
uses. Because of (1) the cost of storing 2°U and (2) arms control considerations, the U.S. government
must decide how much of the existing >’U inventory should be kept for future use and how much should
be disposed of as waste. The objective of this report is to provide technical and economic input to make

a use-or-dispose decision.

ES1.1 INVENTORY

Approximately 2 tons of **U are in inventory. About 1 ton of it is in the form of separated *°U, and
a similar quantity is in the form of spent nuclear fuel (SNF) (Table ES.1). The SNF ***U contains
multiple uranium i1sotopic impurities and fission products. The fission products can be removed by
chemical processing; however, the uranium impurities can not be removed. The SNF #**U is of a lower
quality and is not further discussed herein. Special target fabrication, reactor irradiation techniques, and
aqueous separations techniques are required to produce high-quality 2°U. Much of the separated ***U in
the current inventory was produced using these techniques. Some of this material is relatively pure U,
while the rest contains various uranium isotopic impurities which limit its use. Therefore, it is possible
to have both a shortage of high-quality *°U and a surplus of low-quality **U. A decision about what
material to keep and what to dispose of must be made on a category-by-category basis.

The inventory contains 2°U with both uranium isotopic and chemical impurities. The costs to
produce isotopically pure **U are orders of magnitude greater than those associated with removing
chemical impurities from the uranium. Consequently, the inventory is categorized by the isotopic

composition of different batches of 2°U. The inventory has been divided into three categories.

). 4%
Table ES.1. Quality of major batches of ***U in inventory

 

 

 

 

 

Uranium isotopics Uses
Type Total U N 2 ]
(kg) U U Medical
(kg) (ppm) **Th (g) Other
Uranium-233 in separated form*
High isotopic quality 627.6 607.7 <15 23.9° Yes
Intermediate 1sotopic quality 108.0 95.5 >100 8.7 Yes
Low isotopic quality 1085.2 102.0 125 No
Total 1820.8 805.3 451
Uranium-233 in SNF and targets
High isotopic quality 0.0 0.0 0.0 Yes
Intermediate 1sotopic quality 523.7 501.0 220 0.0° Yes
Low isotopic quality 2528.4 _403.7 0.0° No
Total 3052.1 904.7 0.0

 

“There are three major fissile materials: *°U, **Pu, and **U. The United States has in excess of 100 tons of separated ***Pu
and in excess of 500 tons of separated ***U (HEU). The inventory of separated **U is <1 ton.

% About half of the high-quality separated >*’U and all the SNF is mixed with thorium which prevents practical near-term
recovery of medical 1sotopes. Ifit is desired to produce medical inventories from this inventory, the thorium must be separated,
the 2°U aged for several years for decay product buildup, and then the recovery of medical isotopes can be initiated.

ES1.1.1 High-Quality *°U

High-quality *’U contains only small quantities of other uranium isotopes. About half of this
inventory is in the form of relatively chemically pure oxides. Most of the remaining inventory is **UQ,

mixed with thorium oxide. The thorium oxide can be chemically separated from the uranium.
ES1.1.2 Intermediate-Quality >*U

Intermediate-quality ***U has a significant radiation field associated with it that necessitates remote
handling of this material. It contains significant quantities of the impurity uranium-232 (**U).
Uranium-232 decays to thallium-208, which, in turn, decays and emits a 2.6-MeV gamma ray. For high
concentrations of 22U (140 ppm), the radiation field for a typical package at secular equilibrium
approaches 25 R/h at 1 ft. For many applications, intermediate-quality *°U can not be used because of

the heavy shielding required for worker protection.

XVl
ES1.1.3 Low-Quality 2*U

Low-quality 2°U contains large quantities of other uranium isotopes. Almost all of this inventory is
the Consolidated Edison Uranium Solidification Program ***U, which is about half the total uranium (12
wt % of the 2°U) in the total separated ***U inventory, has an isotopic composition of ~10% ***U and
~76% uranium-235 (**°U), and has a high radiation field because of the *?U content. There are very
limited possible uses of this °U. There are hundreds of tons of HEU thus, this inventory is not a

significant sources of *°U.

ES1.2 PRODUCTION AND STORAGE COSTS

It is estimated that the original production costs of high-quality 2*°U were $2 to 4 million/kg. Low-
quality material is much less expensive since it can be produced in a light-water reactor (LWR).
Irradiation service costs (excluding target fabrication and chemical separation costs) to produce ***U
today using the Advanced Test Reactor (ATR) in Idaho are estimated at ~$30 million/kg. Because of the
shutdown of facilities, the U.S. Department of Energy (DOE) production capability is limited. The ATR,
which 1s the largest DOE reactor currently operating, could produce only ~0.3 kg/year. Only India has a
current capability to produce significant quantities of high-quality 2°U. Newer production techniques
using heavy-water reactors may lower this cost.

Current storage costs are significant. Long-term facility costs and short-term transient costs,
associated with Defense Nuclear Facilities Safety Board recommendation 97-1, total ~$10 million/year.
Long-term storage costs, after current storage issues are resolved may have an incremental storage cost

on the order of a $1 million/year.

ES1.3 USES

Five uses for 2°U have been identified (Table ES.2). The first three uses require relatively small
amounts of material (<100 kg). The other two applications control the size of the long-term need for
233

U.

ES1.3.1 Analytical Chemistry Methods

Uranium-233 is used as a spike (calibration) material in isotopic-dilution mass spectrometry
procedures for the precise determination of uranium inventories and isotopics. These procedures are
commonly used safeguards procedures by the International Atomic Energy Agency (IAEA). The
quantity per analysis is typically a fraction of a milligram. Only high-quality ***U (<10 ppm ***U) with a

minimum of other uranium isotopes is used for this application.

XVil
Table ES.2. Uranium-233 uses and applicable ***U categories

 

 

 

Isotopic quality
Use
High Medium Low
Medical (cancer treatment) Yes” Yes Maybe
Space (deep-space reactor) Yes No No
Analytical (safeguards etc.) Yes No No
Weapons (test, use) Yes No No
Nuclear fuel cycle research and development Yes Yes No

(proliferation resistant fuel cycles)

 

“About half of the high quality ***U is mixed with thorium which prevents practical near-term recovery of medical isotopes. If
it is desired to produce medical inventories from this inventory, the thorium must be separated, the *°U aged for several years for
decay product buildup, and then the recovery of medical 1sotopes can be initiated.

ES1.3.2 Weapons Tests

Uranium-233 has been used historically as an analytical tracer in weapons tests and may again be
used 1n this application if weapons tests are resumed. It is also a weapons-usable material. The IJAEA
has defined a significant quantity of >**U (the amount necessary for a nuclear weapon) as 8 kg. If the
United States were to choose to develop nuclear weapons using >°U, some multiple of 8 kg would be
needed for weapons development and testing until large **U production systems were put into operation.

Only high-isotopic-quality ***U (<20 ppm **?U) would be used for any weapons application.
ES1.3.3 Minimum Mass Reactors (Space and Other Special-Purpose Reactors)

Over a limited range of power demands, ***U (because of its nuclear properties) can be used to build
minimum-mass, nuclear reactors. Such nuclear reactor characteristics are desired for certain special
missions such as deep-space, power-producing reactor systems where there are extreme economic
penalties associated with extra weight.

When considering minimum-mass power systems as a function of power output, 2°U reactors are the
minimum mass systems between small 1sotopic power sources (such as plutonium-238 heat sources) and
larger reactors using HEU (for which the total energy demand controls the fissile inventory of the reactor
as opposed to the critical mass). The future market for nuclear reactors in this narrow range of power
demand is unknown. Only high-quality 2°U (<10 ppm ***U) would be used for this application to

minimize shielding weight before reactor startup.

XViil
ES1.3.4 Medical Applications

Clinical trials are underway using bismuth-213 (**’Bi), a secondary decay product of ***U, to treat
cancer. The preliminary results are encouraging. If the trials are successful, **U would become the
initial source of *"*Bi for medical applications. The *"’Bi from DOE ***U inventories would be sufficient
such as to treat ~100,000 patients per year; however, if ?°Bi became the preferred treatment option for
several cancers, additional methods to produce *"*Bi would be required to meet the demand.

There are multiple methods to produce *"*Bi. The only deployed method is recovery of thorium-229
(**Th), the first decay product of ***U, from the ***U and the subsequent decay of the separated ***Th to
2BBi. The **°Th has a half-life of 7,340 years. The half life of ?*Bi is ~46 min. Consequently, the
extracted **Th provides a long-term source of *’Bi. It requires about 10 years to build up sufficient
22Th in #°U such as to make it practical to recover new *°Th from the ?°U; thus, the *’U can be
effectively processed for recovery of the ***Th only about once a decade. The cost of producing 2"*Bi via
other routes 1s not well defined.

Thorium-229 can be recovered from most of the **°U inventory, but there are limitations for recovery
of ?Th from some of the ***U inventory because of chemical impurities. There are several organizations
examining alternative production techniques. Ongoing economic studies within the next 2 years may be

able to determine if 2°U has a long-term (multi-decade) value as a source of 2"’Bi.
ES1.3.5 Power Reactors

There is one naturally occurring fissile material (*°U), and there are two natural fertile materials
[uranium-238 (***U) and thorium-232] that can produce fissile materials (respectively, 2°Pu and #°U).
Consequently, nuclear reactor fuel cycles are either uranium—plutonium, thorium—>**U, or combination
fuel cycles. With the exception of a small 2*U—fueled research reactor in India, all nuclear reactors today
use some type of urantum—plutonium fuel cycle. LWRs are today the dominant type of nuclear power
reactor. They use low-enriched uranium (3—5% *°U in #**U), which produces plutonium, some of which
1s burned as fuel.

Once-through and breeder fuel cycles exist that use thorium and 2°U. There are several potential

advantages of such fuel cycles:

»  Proliferation-resistant fuel cycles. Uranium-233 once-through and recycle reactor fuel cycles are
much more proliferation resistant than uranium-plutonium fuel cycles. These advanced fuel
cycles produce little or no plutonium. The ***U is (1) either isotopically diluted with 2*U so that
it can not be used in nuclear weapons or (2) created in a fuel cycle that is designed to produce
large quantities of 22U with the 2*U. As a consequence of the 22U content, the recycled uranium
with a high gamma radiation field would be self-safeguarded and would require remote handling,

X1X
*  Reduced fuel consumption in LWRs. High-burnup, once-through LWR fuel cycles that use
thorium and breed *°U may require less uranium than do once-through conventional fuel cycles.
With the recent development of higher-burnup fuels, there may become an economic incentive to
deploy once-through fuel cycles using a combination of thorium and uranium. This is an active
current area of research.

«  Superior waste form. Thorium—>U fuel cycles produce SNF that has a high thorium content.
The performance of thortum-containing SNF in a geological repository is generally better than
that of uranium SNF because thorium fuels are chemically more stable.

e Resource abundance. Thorium is about four times more abundant than uranium.

Thorium fuel cycles, which all generate 2°U, have been investigated but never deployed. In part, this
is an historical accident which saw the early development (ahead of 2°U) of uranium-plutonium
technologies for national defense. In part, this is a consequence of economics and technology. Recent
technical developments and renewed interests in proliferation-resistant fuel cycles have resulted in
increased research on thorium—>>*U reactor fuel cycles in Europe, Japan, India, Russia, Canada, and the
United States.

There are also ongoing investigations of thorium—>**U fuel cycles for nonreactor electric power
systems using accelerators (energy amplifiers). In these systems, spallation accelerators produce energy
by spallation of heavy atoms. Many of these advanced power concepts also propose using variants of
thorium—>**U fuel cycles for the same reasons that they have been considered for nuclear reactors.

With respect to using the existing >**U inventory for development of thorium—>°U fuel cycles, the
question is whether the United States wants to maintain the option to conduct development programs on
thorium->°U fuel cycles—including the options to develop proliferation-resistant fuel cycles. The nation
retains the option to rapidly and efficiently develop any plutonium fuel cycle due to the inventories of
plutonium (tens of tons) being maintained for the weapons program. The option for development of
thorium—>>*U fuel cycles requires that much of the smaller ***U inventory be kept. For these applications,
relatively pure 2°U is needed to provide experimental data without the complications of impure
materials. For such applications, 500 to 1,500 kg of high-isotopic quality ***U is required. This implies
that the entire inventory of high-isotopic-quality 2**U (627.6 kg) and preferably all the intermediate-
isotopic quality *°U (92.5 kg) should be kept. The low-isotopic-quality ***U (half of the separated *°U
inventory in terms of uranium with ~12% of the **U) would be of limited or no value for this

application.

XX
ES1.4 CONCLUSIONS

The cost of replacing the existing inventory of clean 2°U would be many billions of dollars using
current systems and would require centuries to replace with existing capabilities. Consequently,
decisions concerning what material to keep and what material to dispose of should be made with care.
The quantities of *U that should be kept for potential future use are controlled by three questions: What
is the need for decay products from *°U for medical applications? Does the United States want to
maintain the capability to investigate >*U—thorium fuel cycles—including proliferation-resistant fuel
cycles? Are there unidentified uses for 2?U? All other potential uses would require saving <100 kg of
23U for future uses.

Except for possible near-term medical applications, the low-isotopic-quality >**U has little or no
future value. This includes ~100 kg of 2°U (~12% of the separated *°U) and over one-half the total
uranium in the separated 2’U inventory. The cost of recovering medical isotopes from this material will
be an order of magnitude higher than from other sources because this material is about 10% ***U diluted

in HEU.

XX1
ABSTRACT

The United States produced a significant quantity of uranium-233 (**U) during the cold war in
support of national defense and other missions. An evaluation was made to determine what quantities of
233U should be kept for potential uses under various sets of assumptions. There are significant storage
costs for 2°U; however, it would cost many billions of dollars to replace this °U. There are limited
current uses of 2°U, but there are significant potential future uses. The quantities of 2*°U that should be
kept for potential use are controlled by three questions: What is the need for decay products from ***U
for medical applications such as cancer treatment? Does the United States want to maintain the
capability to investigate **U—thorium fuel cycles—including the options for development of
proliferation-resistant nuclear fuel cycles? Are there unidentified uses for 2°U? All other potential uses
would require saving <100 kg of U for future uses. Under most scenarios, the high— and intermediate-
isotopic-quality *°U (703.2 kg **U in 735.6 kg of uranium) is kept, and the low-isotopic-quality material
(102 kg U in 1085.2 kg of uranium) is disposed of.

XX111
1. INTRODUCTION

1.1 OBJECTIVES

Since the end of the cold war, the United States has been evaluating what weapons-usable fissile
materials should be kept for future uses and what fissile materials should be disposed of. There are two

reasons to dispose of excess fissile materials.

» Arms control. The United States has initiated a program with Russia to reduce inventories of
weapons-usable fissile materials. This effort is to mitigate the risks of nuclear war between
weapons states and the risks from theft of weapons-usable materials by third parties.

» Storage costs. The costs of preparing and storing weapons-usable materials is significant.
Consequently, there are economic incentives for disposal of excess material.

However, the cost to produce fissile materials is very high. There are potential future uses.
Consequently, there 1s a trade-off between keeping fissile materials for possible future use and disposing
of those fissile materials.

Weapons-usable fissile materials include plutonium, high-enriched uranium (HEU), and uranium-233
(**U). National decisions have been made concerning what plutonium and HEU to dispose of, and what
to keep. No decisions have been made on what >°U to dispose of and what **°U to keep.

The objectives of this report is to characterize the *°U inventory, define potential uses for 2°U, and
determine what 2°U should be kept using different sets of assumptions. It is to provide the technical

basis for future decisions on what 2*U should be kept for future needs.

1.2 CHARACTERISTICS OF *°U USE OR DISPOSE DECISIONS

Uranium-233 can be converted from weapons-usable to non-weapons-usable 2°U by isotopically
diluting the #°U with ***U to a concentration that is <12 wt % ***U. Isotopic dilution meets the goal of
arms control. Isotopically diluted ***U can be used for some (but not all) potential applications.
Consequently, there are two fundamental »**U use-or-dispose decisions: (1) What pure **°U should be

kept for future uses? and (2) What isotopically diluted >**U should be kept for future uses?

1.3 CAVEATS

This report offers no recommendations on the preferred use-or-dispose decision. However, it does

recommend what portions of the 2°U inventory should be kept under different sets of assumptions.
1.4 ORGANIZATION OF THIS REPORT

Section 2 summarizes U characteristics and provides the inventory data required. These include
the isotopic and chemical impurities in different batches of °U. Section 3 estimates the production costs
for #°U and, thus, the potential economic penalties if too much #°U is disposed. Section 4 describes
existing and future uses for °U. This narrative includes how much ***U should be kept for each use with

different assumptions and what material in inventory would be useful for each application.
2. CHARACTERISTICS AND INVENTORY OF *°U

The inventory of ***U contains materials with different isotopic and chemical compositions.
Accordingly, the value of these materials for different purposes varies widely. The characteristics and

inventory properties are summarized herein.

2.1 CHARACTERISTICS

Table 2.1 shows the characteristics of 2°U as compared to those of the other two weapons-usable

materials—weapons-grade plutonium (WGP) and HEU.

Table 2.1. Characteristics of weapons-usable materials

 

Fissile material

 

 

Characteristic
Plutonium HEU 2y
Production Neutron bombardment of Separation from natural ~ Neutron bombardment of
28U uranium thorium-232 (***Th)
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”
Chemuical 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 % 2**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 non-weapons-usable ***U.
"Isotopic dilution of ***U with **U to this limit minimizes the potential for nuclear criticality in disposal facilities.
2.1.1 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 2°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 2°U are similar to those for WGP (Fig. 2.1).

In the production of #°U, some uranium-232 (**?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 stable lead (***Pb) and produces a high-energy, 2.6-MeV gamma ray. The
concentration of 22U determines the radiation shielding required to protect workers. Superior-quality
233U contains very low levels [~1 part per million (ppm)] of 22U and has correspondingly low levels of
gamma radiation. Low-quality 2*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 (see Appendix A).

The ***U in low-quality 2°U also impacts the requirements of off-gas systems for processing these
materials. Uranium-232 decays through several isotopes to the noble gas radon-220 (**°Rn), which
decays further 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 may require (1) a HEPA filter to collect solids
(including the precursors to **°Rn), (2) charcoal beds, delay lines (~10 min), or other special equipment
to hold the radon in the off-gas system that goes through the first HEPA filter until the **°Rn decays to a
solid material, and (3) a second HEPA filter to remove the solid *°Rn decay products. Typical off-gas
systems designed for HEU or plutonium are not acceptable for °U with a high *?U content because they
do not contain the double HEPA filters with the time delay between the HEPA filters required to avoid
release of *°Rn to the environment.

There is an important radiochemical characteristic of this system. If *U is chemically purified by
removing 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
occurring workers. It takes time (days to weeks) for the radioactive 2*?U decay products that emit gamma
rays to build up to high enough concentrations such that thick radiation shielding 1s required to protect
the workers. Very clean processing systems are required for this type of operation. If **U contamination
1s allowed to remain in the system, radiation levels will build up with time and can dominate the radiation
field from such processes. The buildup and decay of 2°U, #**U, and decay products are shown in Fig. 2.2
for 2°U with high concentrations of #**U. The first set of peaks are from the buildup and subsequent
decrease of the decay products of 2?U. The second set of peaks are from the buildup and subsequent
decrease of the decay products of 2*U. The curve for gamma-ray generation vs time since purification of

the uranium shows that, for a relatively short time after purification, the gamma-radiation doses are low.
ORNL DWG 98C-488R

 

 

 

Alpha Activity (Relative) = 1
No Significant Gamma

 

 

 

 

 

Alpha Activity (Relative) ~10*
Soft Gamma

- Minimal Shielding

- Leaded Gloves Acceptable

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 
   
 

 

 

 

 

Alpha Activity (Relative) ~10°

Hard Gamma (2.6 Mev)

- Time Variable

- Gamma From Decay
Product of 232U Impurity

- Clean 233U Processed in

Glove Box Hot Cell Glove Box

 

 

 

 

 

 

Fig. 2.1. Different fissile materials require varying handling procedures.
Gamma Exposure Rate at 1 ft (R/h)

s
N

—lb
<

o

ORNL DWG 98C-157R3

 

 

 

 

Y (2.6 MeV)

/

232 w .~ 220, . . 2087 . . 208py,
(T, =72years) (T,,=56s) (T, = Smin) (Stable) d
/ 233

  
  

Uman»maa»man—

| ] i
100 10,000
Time (years)

Fig. 2.2. Gamma exposure for 1 kg of ***U with 100 ppm of >**U.

 

209
(T, P 160,000 years) (Stable)

Bi

 

 

1,000,000
Uranium-233 with high-concentrations of 2**U has much higher handling costs associated with it.

Consequently, the 22U content becomes an important measure of the quality of 2°U.
2.1.2 Nuclear Criticality

The nuclear characteristics of 2°U are significantly different from those of plutonium or HEU. The
subcritical, single-parameter, mass limit of 2°U is about 520 g (Forsberg, January—March 1997,
American Nuclear Society 1981). This is significantly less than that of uranium-235 (*°U) (700 g) and
slightly greater than that of plutonium-239 (**Pu) (450 g). Furthermore, the behavior of >**U in a nuclear
reactor 1s significantly different than that of other fissile materials. Consequently, there are some types

of reactor designs for which **U is the preferred fuel.
2.1.3 Safeguards

Uranium-233 is a weapons-usable material. As a fissile material, 2°U is similar to WGP. The 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 is that quantity of material requiring nuclear weapons-type
security to prevent theft of the materials.

National and international safeguards requirements [U.S. Department of Energy (DOE) orders,

U.S. Nuclear Regulatory Commission regulations, and IAEA agreements] for weapons-usable materials
have been developed for HEU and WGP; however, the requirements are not developed fully for **U. For
uranium containing >°U, these regulatory requirements recognize that only HEU can be made into
nuclear weapons. Natural uranium, depleted uranium (DU), and low-enriched uranium (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 as fuel 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 2°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 2°U to non-weapons-usable material by diluting it with **U is understood, but the regulations
and other institutional agreements are not in place. Activities are underway to obtain institutional
agreements to define the level of isotopic dilution that eliminates the weapons potential of 2°U
(Forsberg et al. March 1998). The isotopic purity that renders 2°U non-weapons-usable [<12 wt % U
in uranium-238 (***U)] is less than that for HEU (<20 wt % *°U in %U).

For mixtures of 2°U, #°U, and ***U, effectively non-weapons-usable uranium is defined by the

following formula;

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

 

<0.12

One kilogram of #°U requires 7.407 kg of DU containing 0.2 wt % ***U to convert *°U to non-
weapons-usable uranium (<12 wt% 2**U in 2*U). If the **U is isotopically diluted to this concentration,

it remains useful for some applications, but not for others.

2.2 INVENTORY

About 2 tons of 2°U are in inventory. About 1 ton is in the form of separated ***U (Table 2.2), and a
similar quantity is in the form of spent nuclear fuel (SNF) (Table 2.3). The SNF *°U contains multiple
uranium isotopic impurities and fission products and thus has limited applications unless it is reprocessed
to remove the highly-radioactive chemical impurities. It is not further discussed herein.

Special target fabrication and reactor irradiation techniques are required to produce high-quality >°U.
It is this material that was reprocessed to produce most of the separated >°U in the current inventory.
Some of this material is relatively pure **U, while other material contains various uranium isotopic
impurities which limit its use. Therefore, it is possible to have both a shortage of high-quality *°U and a
surplus of low-quality °U. A decision about what material to keep and to dispose of must be made on a
category-by-category basis.

The inventory contains 2*°U with uranium isotopic and chemical impurities. The cost to produce
isotopically pure U is orders of magnitude greater than the cost associated with chemical purification
of uranium. Consequently, the inventory is categorized by the isotopic composition of different batches

of 2*U. There are two types of isotopic impurities.
Table 2.2. Quality of major batches of ***U in inventory®

 

Uranium isotopics

 

 

Batch . . . . . Total U
;{: Location/designation Material and packaging (()kag) 2337 25 232q5
(kg) (kg) (ppm)*
High 1sotopic quality with limited chemical impurities
1 Oak Ridge National U;0O4 monolith in 27 welded 65.2 60.3 0.0 15
Laboratory (ORNL) stainless steel cans
2 ORNL (2 similar batches) UQ, powder in 247 stainless 108.8 103.1 0.0 4-9
steel screw-top cans
3 ORNL U,;0O4 powder 1n 1,645 welded 46.0 45 0.0 6
stainless steel plates
4 Multiple/Remaining small lots Many forms and packages 49.0 479 ~0.0
Subtotal 269.0 256.3
High 1sotopic quality with chemical diluents (ThO, or ZrO,)
5aӢ  Idaho National Environmental ~ Unirradiated rods and pellets 29.5 28.5 0.0 9
and Engineering Laboratory with 483 kg ThO,
(INEEL)/Light-Water Breeder
Reactor (LWBR)
5b INEEL/LWBR (ZrO,) Unirradiated rods and pellets 5.6 55 0.0 38
made of **UQ, and ZrO,
6° INEEL/LWBR Unirradiated LWBR fuel with 323.5 317.4 0.0 9
14 t natural thortum
Subtotal 358.6 3514
Intermediate 1sotopic quality
7 ORNL/Savannah River Site UO,; powder in 140 welded 67.4 61.6 0.0 156
(SRS) mner aluminum cans
8 ORNL/Molten Salt Reactor UO, powder after conversion 40.6 33.9 160-200
Experiment (MSRE)?
Subtotal 108.0 95.5
Low isotopic quality
9 ORNL/Consolidated Edison U,;0O¢ monolith in 403 welded 1042.6 101.1  796.3 120
Uranium Solidification stainless steel CEUSP cans
Program (CEUSP)
10 Clean/Y-12 UO, powder in 5 cans 42.6 0.9 387 6
Subtotal 1085.2 102
Total 1820.8 805.3

 

“Based on the uranium content.

’The uranium is in the form of UO,~ThO, fuel pellets with 1 to 10 wt % ***U. The average assay is ~2 wt % 2*3U.

‘One drum of 188 g ***U metal in 9.3 kg thorium metal.
‘Material in inventory and being recovered from the MSRE. The material will be converted to oxide form for storage.
Table 2.3. Quality of major batches of >**U in SNF*

 

Uranium isotopics

 

 

 

Batch No. Site (reactor) T(()lt(agl)U 23 0
(kg) (ppm)
Intermediate 1sotopic quality
1 INEEL (LWBR) 523.7 501.0 220.
Low 1sotopic quality
2 SRS (Dresden) 684.0 154 High
3 SRS (Elk River) 2243 14.7 High
4 SRS (Sodium Reactor Experiment) 154.9 1.1 High
5 INEEL (Ft. St. Vrain) 308.3 90.1 48.3
6 Colorado (Ft. St. Vrain) 822.5 236.0 53.4
7 INEEL (Peach Bottom I) 206.6 20.5 7.1
8 INEEL (Peach Bottom II) 127.8 _259 58.6
Totals 3052.1 904.7

 

“No high-isotopic-quality ***U is in the current SNF inventory.

Uranium-232. Uranium-232 determines handling practices. This isotope decays to 2°*TI, which,
in turn, decays and emits a 2.6-MeV gamma ray. Uranium-233 with high levels of ***U has a
significant gamma radiation field which necessitates expensive RH of this material and which, in
turn, limits its potential uses. For separated *°U with low levels of ***U, the uranium can be
purified and handled for several weeks or months before the decay products with high gamma-
radiation fields increase to a level such that the RH of the **U is required. For most
applications, 2*U with low concentrations of *?U is required.

Other uranium isotopes. Two batches of separated ***U contain large quantities of 2°U. The
CEUSP ***U, which is about half the total uranium (12% of the **°U) in the separated *°U
inventory, has an isotopic composition of ~10% ***U and ~76% ***U. The Y-12 ***U contains
only a few percent 2°U in #°U. There are limited possible uses of this material as 2°U. There
are hundreds of tons of HEU; thus, these inventories are not significant sources of *°U.

Based on the previous considerations, the separated inventory can be divided into four major

categories: high isotopic quality with limited chemical impurities, high 1sotopic quality with chemical

diluents, intermediate isotopic quality, and low isotopic quality. The inventory contains ~1,800 kg of

10
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.
2.2.1 High-Isotopic-Quality **U with Limited Chemical Impurities

High-isotopic-quality 2°U contains low concentrations of **U and other uranium isotopic impurities

and few chemical impurities.
2.2.2 High-Isotopic-Quality ***U with Chemical Diluents

High-isotopic-quality ***U with chemical diluents is similar to the high-isotopic-quality 2**U—except
for the presence of one or more other elements that could be separated from the ***U by chemical
processes. Almost all of this inventory is from the LWBR program, which investigated the use of
thorium—>*U nuclear fuels. When the program was shut down, one unirradiated fresh fuel assembly, fuel
rods, fuel pellets, and other assorted materials were placed in storage at INEEL. There are several
batches. All of this material is high-quality **U with a low **U content. Because it was to become
reactor fuel, the **U was mixed with thorium or zirconium oxides; thus, it has been chemically diluted
and 1s not in a pure chemical form. Chemical separations would likely be required before this material
could be used.

While the material is in several types of packages, it primarily consists of 1 to 12 wt % *°UQ, in
high-fired (1,750°C for 12 h) ThO,. The average assay is ~2.5 wt % **UQ, 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.
2.2.3 Intermediate-Isotopic-Quality *°U

Intermediate isotopic quality 2°U contains a significant **U content and, thus, has a significant
radiation field associated with it. The intermediate-isotopic-quality inventory consists of two batches of
material stored at ORNL. The **°U originally produced at SRS is a chemically pure oxide. The MSRE
233U is partly in storage and partly in the MSRE reactor salt solution. The *°U is currently being
separated from this salt to address safety issues identified in the Defense Nuclear Facility Safety Board

recommendation 94-1. For safe storage, the material will be purified and converted to an oxide.

11
2.2.4 Low-Isotopic-Quality >*U

The low-isotopic-quality 2°U contains high concentrations (tens of percent) of other uranium
isotopes. There is one large batch of low-quality U (CEUSP ***U) that consists of half of the inventory,
as measured by the uranium content. The CEUSP material is a mixture of ~10 wt % U, ~76 wt % *°U,
and other uranium isotopes. It is *°U isotopically diluted with HEU. The CEUSP *°U also has a high
concentration of 2?U. This results in a significant gamma radiation field near the containers. The
CEUSP *°U contains both cadmium and gadolinium oxides that were added for criticality control.

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 at the
West Valley commercial fuel processing facility with the **U shipped in the form of a uranyl-nitrate
aqueous solution to ORNL, where it was solidified for storage. Because all of this material was stored as

a liquid solution 1n a single tank, it is a single, homogeneous batch of material.

12
3. PRODUCTION COSTS FOR **U

3.1 HISTORIC PRODUCTION COSTS

No detailed production cost records for 2°U have been identified. Much of the cost is associated
with the operation of large production reactors and reprocessing plants that co-produced plutonium,
tritium, >**U and other products. Rough estimates can be made. The production techniques of high-grade
23U are similar to those of WGP (Orth 1979). Plutonium is produced by irradiating ***U in a production
reactor, whereas 2°U is produced by irradiating > Th in a production reactor. In both cases, complex
chemical separations are required in shielded facilities. A rough estimate of production costs for 2°U can
be made by assuming that the costs for °U and plutonium are similar.

Recent studies have begun to evaluate the costs of the cold war and the costs associated with nuclear
weapons deployment during the cold war (Schwartz 1998). These studies provide one basis for
estimating historical costs. The United States has declared that it produced ~100 tons of plutonium
during the cold war. Most of DOE's cold-war cleanup costs are from the production and purification of
this plutonium. The cleanup costs are estimated at $300 billion. The cost to produce weapons materials
(primarily plutonium and HEU) are estimated at somewhat under $200 billion. This suggests that the
costs to produce WGP were $2 to 4 x 10%kg. Similar costs would be expected for the production of
high-isotopic-quality **U. Low-isotopic-quality material is much less expensive because it can be

produced in a light-water reactor (LWR).

3.2 CURRENT SALES PRICE AND PRODUCTION COSTS

The United States sells 2**U—primarily for analytical purposes (DOE December 1998). The sales
price is $6.95/mg. This is equivalent to ~$7 x 10%/kg (larger sales may have negotiated prices). This
price partly reflects the small quantities of material and the purity requirements for **U when used for
analytical purposes.

Production costs for 2°U today would be very high because the United States has shut down its large
production reactors. Irradiation costs (excluding target fabrication and chemical separation costs) to
produce *°U today using the Advanced Test Reactor (ATR) in Idaho are estimated at ~$3 x 107/kg. The
ATR, which 1s the largest DOE reactor currently operating, could produce only ~0.3 kg/year.
Worldwide, only India (Ganguly et al. 1991) has a current capability to produce significant quantities of
high-quality ***U.

13
3.3 FUTURE PRODUCTION COSTS

Newer production techniques using heavy-water reactors may lower this cost. Because of historical
factors, the production and use of 2°U were investigated much later than were the production and use of
uranium and plutonium. Most of the research was done in the 1960s and early 1970s. That research
indicated lower-cost production routes, but large-scale research on ***U ceased before any of those

production methods could be developed.

14
4. USES OF *°U

There are several potential uses for 2°U and its decay products. Figure 4.1 summarizes the larger
potential uses. By definition, only known uses of 2°U are described herein. There is no assurance that

all of the potentially significant uses of **U have been identified.

4.1 MEDICAL APPLICATIONS
4.1.1 Use

One potential large-scale use for 2°U involves one of its decay products, bismuth-213 (*"*Bi) for
cancer treatment (Table 4.1). Specifically of interest is the use of antitumor antibodies radiolabled with
an alpha emitter (Knapp and Mirzadeh 1994; Geerlings 1993). In this therapy, the radioisotope, *"’Bi, is
attached to antibodies that, in turn, attach to cancer cells; the resulting alpha emissions kill these cells
with high efficiency. Initial clinical trials using *’Bi on human patients at the Memorial Sloan-Kettering
Cancer Center Hospital in New York City have been favorable.

The goal of radiotherapy is to kill the cancer cells without killing healthy cells and the patient. The
interest in 2°Bi, as compared to other radioisotopes, is that its nuclear characteristics may maximize
damage to cancer cells while minimizing damage to healthy cells. This characteristic allows higher
concentrations of the radioisotope to more effectively kill cancer cells without killing the patient from

radiation or causing excess radiation exposure to other persons.

*  High local damage. Radiation therapy has long been used to treat cancer. Alpha emitters
compared to other radiation sources (x-ray, gamma, beta, etc.) deposit most of their energy in a
very small volume within a few cell diameters. The large local energy deposition provides a
higher assurance that the specific cell is destroyed, not just damaged. It is estimated that only
two 2Bi decays will kill a cancer cell.

*  Auxiliary damage control. In most types of radiation therapy, the radiation is concentrated on
cancer cells, but healthy cells also receive high radiation doses. For example, if x-rays are used,
many of the x-rays will be absorbed into healthy cells. Because alpha damage 1s very localized,
secondary damage is minimized. This outcome is particularly important in treatment of certain
cancers (e.g., leukemia) and other diseases (e.g., meningitis) where single cells or small clusters
of cells are the targets that are interdispersed among healthy cells. Conventional radiation
therapy will kill large numbers of healthy cells and have the potential to harm the patient
(Feinendegen 1996).

*  Minimal long-term damage. Many alpha emitters could be used for medical applications.
Unfortunately, most alpha emitters decay through many additional decays to a stable 1sotope.
Each of these subsequent decays creates radiation damage beyond the cancerous cell that was
destroyed. These longer-term effects can adversely impact the health of both patients and
doctors by several mechanisms.

15
 

 

 

 

 

ORNL DWG 97C-149R5

Medical
(Cancer Treatment)

 

_ Low-Mass, Deep-Space
” Nuclear Power Source

 

)

Analytical
Measurements
(For safeguards,
safety, etc.)

     
 

». Weapons Testing/

" Weapons

Development of
» Nuclear Power
" Nonproliferation
Fuel Cycles

 

Fig. 4.1. Potential uses for **U

16
Table 4.1. Summary: medical applications

 

Application Treat cancer patients with 2°Bi, a decay product of 2*U, to selectively destroy cancer cells in
the body.
Acceptable ***U feeds Near term: all except LWBR ***U; long-term: all

Isotopic purity requirements ~ None. Choice of feed material 1s determined by economics.

Demand

Clinical trials are underway to determine the benefits of using ***Bi for cancer treatment. If
the clinical trials are successful, the entire ***U inventory could be used as a *’Bi source.
There are competing (but not developed) methods to produce °Bi. With high-use scenarios,
alternative production methods would be required to meet demand.

Use implications Nonconsumptive use of 2*U. The desired product is ***Bi-a decay product of ***U.

Technical description The *"*Bi is obtained from #**U by a multistep separation process. The ***U decays to

Assessment

thorium-229 (**Th), which has a 7,340-year half-life. The ***Th is separated from the ***U
and is then used as a source for the short-lived *°Bi.

The LWBR **°U is useful only if the ***U is separated from the natural thorium that is an
integral part of the fuel and if **Th is allowed to build in. The low-isotopic-quality (CEUSP)
30U is a source of 2’Bi, but the cost of 2’Bi recovery from this **U is significantly higher
because of the low concentration of >>*U, and, hence, **Th in this feed stock.

This may be a major use of >*U if ongoing clinical trials on the benefits of *"*Bi for cancer
treatment are successful. Initial clinical trials have been successful. It is the only current
source of ’Bi, but there are alternative production techniques. The lowest-cost *’Bi is from
aged, clean, high-isotopic-quality ***U. Production costs from other **U inventories will be
significantly higher.

 

Lifetime. The objective of a radiopharmaceutical is to cure the patient. If there are long half-
life 1sotopes associated with a particular treatment, these isotopes result in a damaging long-
term radiation dose to the patient and potentially to nearby individuals after treatment.
Bismuth-213 has the desirable characteristic in that it and its decay products all have short
half-lives (Fig 4.2) and quickly decay after destroying the cancer. The half-life of *"’Bi is

45 min. It decays to a stable isotope of bismuth through two additional radioisotopes, one
with a half-life measured in microseconds and the other with a half-life of 3.31 h. There are
no long-lived radioisotopes to cause future damage to the patient or nearby individuals.

Secondary radiation doses. Many alpha emitters decay through other radionuclides that emit
high doses of radiation. For example, *'*Bi has been used in radiation therapy. It is an alpha
emitter like ?"’Bi, but it decays to *°*Tl that, in turn, emits a 2.6-MeV gamma ray. This
gamma ray irradiates both the patient and the medical staff. In contrast, ?°Bi primarily
decays by beta emission to **Po, which, in turn, decays to **’Pb by alpha emission in 4 x 107
s. Both decays are simultaneous in terms of the destruction of cancer cells. The **Pb, in
turn, decays with a low-energy beta ray to stable **Bi.

17
ORNL DWG 99C-280

 

 

 

233U
1.59x 10°y
ol
481 MeV
229Th
7340y
ol
4.86 MeV
225AC o8 225Ra
10d 0.11 MeV 1484
ol
5.75 MeV
221Fr
48m
ol
6.36 MeV
32 ms
ol
7.07 MeV
213PO 255 213Bi (46 m)
4 us 0.46 MeV
| o o (2.2%)
| 837 MeV 0.13 MeV

 

209Bi ZOPb 209Tl
Stable ~ 0.20MeV  331h 276 MeV 22m

 

Fig. 4.2 *°U decay chain.

18
4.1.2 Production Methods for ?'*Bi

There are many potential methods that may be used to produce *°Bi (Mirzadeh 1998). The
production of 2*Bi from #**U is the only current production method; however, other production methods

are being investigated.
4.1.2.1 Production from U
4.1.2.1.1 Production Method

Bismuth is a decay product in the *°U decay chain with a half-life of 46 min. A three-step process is
required to recover the 2°Bi for medical use from **U (Fig. 4.3). The decay product, **Th, is recovered
from the U, the decay product actinium-225 (***Ac) is recovered from the thorium, and the decay

product *’Bi is recovered from the **Ac.

o Thorium-229 buildup and separation from *°U. Thorium-229 is the decay product of ***U.
Because of its long half-life (7', = 7,340 years), it slowly builds up in ***U inventories. After
30 years of storage, 1 kg of 2°U will contain ~120 mg of **Th. The separation is accomplished
by first dissolving the *°U in nitric acid. 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. The ***Th is then recovered by washing the ion-exchange
resin with 0.1 M nitric acid. After the thorium has been separated from the uranium, the uranium
1s solidified as an oxide, then aged for several years (~10 years), and the process can then be
repeated to recover fresh *Th.

o Actinium-225 buildup and separation from **Th. The thorium is aged for several weeks to allow
the ingrowth of the decay product ***Ac. Thorium decays to radium-225 (***Ra) (T, = 14.8 d),
and the **’Ra decays to *’Ac (T,, = 10 d). The **’Ac is separated from the ***Th and the other
decay products. Because actinium is not a part of the decay chain of >2U, this separation
removes the undesirable decay product *°*T1 and its precursors. A biomedical generator system
is loaded with *Ac and sent to the hospital. The process can be repeated in several weeks with
the fresh production of actinium from the **Th.

Bismuth-213 buildup and separation from *>Ac. The **’Ac decays to Francium-221 (7, =
4.8 ms), which then decays to *'’At (7, = 32 ms), which next decays to *’Bi. At the hospital,
the 2P°Bi is separated from the *’Ac, converted into the appropriate chemical form, and injected
into the patient. The continuous decay of *’Ac allows the repeated recovery of new *"’Bi from
the *’Ac every day. The ***Ac at the hospital decays completely away. Fresh *’Ac may be
recovered every couple of weeks from the *’Th and sent to the hospital.

The multistep process is carried out in several different locations because of the different process and

facility requirements.

19
ORNL DWG 98C-7582R

 

 

DU
Optional

 

 

 

 

 

\,. Recover Resolidify and Store in
U 2297 py Standardized Packages
lon Exchange for Storage or Disposal

 

 

 

 

 

 

 

 

Safeguards

No Saf;guards 22Th and Other Decay Products
v

 

 

 

29Th T, , = 7340 years

 

Se arate 225AC T1/ = 10 dayS

o2 2
Ac by

lon Exchange

 

   

 

 

 

t

Hot Cell
Hospital

% 225Ac

“Milk” *"*Bi from .
Biomedical Bi
Generator

System 2138 T,,, = 46 min

 

 

   

 

 

 

Fig. 4.3. Flowsheet for *’Bi production from ***U for treatment of cancer.

20
»  Bismuth—actinium separation. The bismuth—actinium separation is done at the hospital. The
half-life of *°Bi is only 46 min. The patient must be near the separation facility so that the *°Bi
does not decay before it is injected into him or her.

o Actinium-thorium separation. The actinlum-thorium separation may be carried out in a hot-cell.
The actinium product has a 10-d half-life; thus, fast shipment to hospitals is required.
Centralized facilities would be used, but there could be more than one facility.

»  Thorium—uranium separation. The initial thorium—uranium separation is usually performed in a
hot cell because of other impurities in the **U (***U). If there is a significant quantity of
weapons-usable 2°U, there are special security requirements to prevent theft of weapons-usable
material. The thorium product is not weapons-usable material and, thus, does not have the
security requirements of the 2°U. The security, hot-cell, and other requirements imply that for
economic reasons, a single facility would likely do all such processing in the United States.
Since the *’Th has a 7,430-year half-life, rapid shipping of the thorium product to the actinium
separation facility is not required.

4.1.2.1.2 Thorium-229 Inventory

All of the #°U in inventory, except the LWBR #**U, can be used in the near-term as a source of *Th
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 were desired to obtain **Th from
the LWBR 2*°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 quantity of **Th in any batch is dependent only upon the quantity of 2*U and the age of the
batch since the **Th was last separated from the uranium. Table 4.2 lists the quantity of **Th in each
batch and the quantity of *’Th per unit of uranium. About 40 g are available from the **°U inventory.
Processing costs are approximately proportional to the quantity of uranium, thus, the lowest cost **’Th

will be recovered from the uranium with the highest concentration of ***Th.
4.1.2.1.3 Production Issues

The option exists to isotopically dilute the 2*U with 2*U to convert it to non-weapons-usable °U.
This eliminates the need for high security in the facility that separates the thorium from the uranium.
Isotopic dilution increases the quantity of uranium to be processed per unit of thorium product, but 1t
reduces security costs. It has been demonstrated that the thorium can be recovered from these more
dilute solutions. However, there has been no economic assessment to determine which option would be
the most economic if the 2*U were to be saved only for medical purposes. The decision to dilute is an

irreversible decision.

21
Table 4.2. Uranium sources of *Th for medical applications

 

 

 

 

Uranium isotopics >®Th Quantities
Batch Location/designation Total U 233 232 229 229
No. (kg) U U Th Th/U
(kg) (ppm) (g (mg/kg)

High 1sotopic quality with limited chemical impurities (Clean)

1 ORNL 65.2 60.3 15 8.2 126

2 ORNL (2 batches) 108.8 103.1 4-9 6.4 59

3 ORNL 46.0 45.0 6 3.7 81

4 Multiple/remaining small lots _49.0 _479 56 114
Subtotal 269.0 256.3 23.9
High 1sotopic quality with chemical diluents (ThO, or ZrO,)

S5a INEEL/LWBR with ThO, 295 28.5 9 a a

5b INEEL/LWBR with ZrO, 5.6 5.5 38

6 INEEL/LWBR with ThO, 3235 317.4 9 a a
Subtotal 358.6 351.4
Intermediate 1sotopic quality

7 ORNL/SRS 67.4 61.6 156 8.7 128

8 ORNL/MSRE/ _40.6 339 >160 _0.0 128
Subtotal 108.0 95.5 8.7
Low isotopic quality

9 ORNL/CEUSP 1042.6 101.1 120 12.5 12

10 Clean/Y-12 _ 426 _09 6 _~0.0 0.0

Subtotal 1085.2 102 12.5
Totals 1820.8 805.3 451

 

“These materials contain natural thorium. Thorium-229 can not be practicable separated from this natural thortum. To use
these materials for production of medical 1sotopes, the materials must be purified (thorium removed) and aged to allow ingrowth
of *Th.

22
If there were no thorium losses when a thorium-actinium separation is conducted, the **Th with a
half-life of 7,340 years could provide actinium for thousands of years. Unfortunately, some of the
thorium “plates-out™” on equipment, and separation processes are not totally efficient. If the separation
process is conducted once a month with an efficiency of 99%, in 100 months only 37% [(0.99)'%°] of the
initial thorium will remain. Losses in the actintum—thorium separation step (not radioactive decay)
determine how often additional **Th must be acquired from ***U. Typical laboratory efficiencies are
about 99.5%; however, in any industrial operation there will be operations that fail and these may control

total long-term losses. Additional development may reduce these losses.
4.1.2.1.4 Current Status

This is the current method to produce 2"°Bi for ongoing research and clinical trials. In the United
States, the primary inventory of 2°U is at ORNL and, thus, ORNL conducts the separations required to
produce **’Ac for the medical community. Several private companies have proposed (under various
conditions and constraints) to DOE to privatize this program and expand the production as needed. A
small production capability exists in Germany for recovery of *’Ac from *Th. There are proposals for

recovery of *’Ac from Russian ***U/**’Th inventories. The size of the Russian inventory is not known.
4.1.2.2 Production from Radium-226 (**Ra)

Several processes are being investigated to produce *°Bi from ***Ra. As a feedstock for the
production of ?"*Bi, ?*Ra has the advantage that it is available in sufficiently large quantities such as to

meet any demand.
4.1.2.2.1 Production Methods

Many organizations are investigating production of ?>Bi from ***Ra—a material that is more
available than 2*U. Most of these organizations are investigating a single production route. The
exception is the European Atomic Energy Community (EURATOM) sponsored work at the Institut fuir
Transurane (ITU) of the Geschellshaft fiir Strahlung and Umweltforschumg, GmbH, Miinchen in
Karlsruhe, where the work has led to a series of European patents pertaining to the production of curie-
quantities of **’Ac and *"’Bi from the irradiation of **°Ra targets, using three approaches. Proposed
production methods include:

»  Gamma-Neutron (y,n) production of **Ac. Actinium-225 can be produced by irradiating ***Ra

with gamma rays to produce **’Ra, which has a half-life of 15 days and decays to *’Ac, which is

then separated from the radium target. The ***Ac is used as a source of ?'°Bi in the same way as
the *’Ac that is produced from **Th.

23
AlphaMed® Inc., a Massachusetts Corporation, is planning commercial production of *"’Bi
generators based on the *°Ra (y,n)-**Ra reaction (Lidsky 1999). AlphaMed® has done proof-of-
principle tests demonstrating production of ***Ra and separation of the desired decay product,

2 Ac, from much larger quantities of radium and its decay products. AlphaMed® has obtained
exclusive licenses to use Massachusetts Institute of Technology (MIT) patents on high-yield
target designs and Eichrom patents for separation technology and generator design. The
measured *’Ac yield is sufficient such as to supply the projected preclinical and clinical trials
(15 Ci/year) for the next 3 years with a single linear accelerator.

The EURATOM program at ITU in Karlsruhe examined gamma-—neutron (y,n) production of
*»>Ra using a reactor. The European patents, EP0752710/LU88637 (Koch, January 8, 1997a),
relates to irradiating *°Ra targets in a flux of epithermal neutrons in a “fast” breeder nuclear
reactor to produce *’Ac by (n,2n) and (y,n) reactions.

Three—neutron (3n) capture production of **Th. Thorium-229 can be produced by irradiating
22°Ra with sequential absorption of three neutrons with two subsequent beta decays. The yield is
~7 mg (~150 mCi)/g of **°Ra irradiated with a thermal-to-epithermal neutron flux ratio of 10 and
a total flux of 1 x 10" neutrons/s-cm* (Mirzadah 1998).

It is estimated (Feinendegen 1998) that about 8.4 g of *Th could be produced per year by
irradiating 100 g of **°Ra in the High Flux Isotope Reactor at ORNL and using existing support
facilities. This estimate assumes an adequate supply of *°Ra. There have been earlier
irradiations of ***Ra to produce **’Ac—an intermediate product on the route to produce **°Th.
These earlier irradiations did not measure the **Th production. Proposals have also been made
by the Russians to produce **Th by this route.

Also, a series of EURATOM patents [US5355394/EP0443497 (Fuger, November 11, 1994) and
EP075210/LU88637 (Van Geel, February 23, 1991)] relate to irradiating **°Ra-targets in a
thermal-neutron reactor at a flux of about 5 x 10'* neutrons/cm?* sec for up to three years in order
to produce **’Ac through thorium-228 (***Th) and **Th decay chains (Fig. 4.2).

Proton—two neutron (p,2n) production of *Ac. The Transuranium Institute in Karlsruhe,
Germany, is investigating the production of *’Ac by bombarding of **°Ra with 20-MeV protons.
In proton irradiation, the *°Ra is converted to the excited state of **’Ac, which then emits two
neutrons forming *’Ac, which is the source of 2"°Bi.

EURATOM patent EP0752709/LU88636 (Koch, January 8, 19975) relates to producing **Ac
from **°Ra targets by bombarding the targets with protons from a cyclotron. This patent does not
specify the conditions. However, some of the boundary conditions are known. The threshold
energies required of these proton interactions with the **°Ra nuclei are between 14 and 16 MeV,
and the upper limit of their energies i1s between 25 and 30 MeV, at which point the protons
penetrate the target without significant interaction. To avoid target melting, the proton-beam
power must be limited; thus, the probable cyclotron beam powers range between 100 pA to 1
mA.

In a personal communication (Koch, June 29, 1999), L. Koch indicated that there were current
plans to produce 1 Ci of ***Ac by this route. A cyclotron, proton accelerator currently used for
generating positron emission tomography scan radioisotopes would irradiate the radium target

24
over a weekend. A dedicated accelerator (Koch, June 30, 1999) could produce as much as 52
Ci/year. No such dedicated facility currently exists, and it would be a significant undertaking to
produce and process *°Ra targets and separate the products at this 52 Ci/year. No costs analyses
are available at this time.

»  Other production techniques. Other production techniques have been identified including
neutron—two-neutron (n,2n) and deuterium—neutron (d,n) nuclear reactions to yield **Ac as the
intermediate product for **Bi production.

4.1.2.2.2 Radium-226 Availability

Radium-226 is currently obtained from inventory. It is estimated that Russia has ~1 kg. About 100 g
of pure material are available from Europe. Significant quantities are in certain waste streams including
~4 kg in waste silos at Fernald, Ohio.

Radium-226 is a decay product in the #*U decay chain. It was recovered from uranium ores in the
late 1800s onward for medical and other purposes. It is the radioisotope used in radium watch dials and
was used on a large scale in World War I for aircraft instrument lighting. It could again be recovered
during the uranium milling process. Because of the limited market, it 1s not currently recovered during
uranium milling operations.

The likely source for additional **Ra would be from Canadian uranium mills. Large quantities are
potentially available. Canada is the world’s largest producer of uranium. Canadian environmental
regulations place strict limits on the quantity of radium dissolved in water from uranium mills.
Consequently, Canadian mills add BaCl, to remove radium from neutralized waste waters
(Sherwood 1983). In high-sulfate solutions, radium is coprecipitated, forming a (Ba, Ra)SO, solid that is

then disposed of. This precipitate would be the raw material from which to obtain purified radium.
4.1.2.2.3 Production Issues

There are multiple options for producing *"’Bi from ***Ra. The primary uncertainty is cost.
Secondary issues include facility availability. Because radium is a highly toxic alpha emitter, it requires

special handling facilities.
4.1.2.3 Production from **Ra

Thorium-229 can be produced by a simple one-step neutron irradiation of ***Ra, which is a natural
radium decay product of *Th—the thorium isotope found in nature. Currently, there is a world surplus
of thortum; thus, there 1s little thorium mining. The U.S. Defense Logistics Agency (Defense National

Stockpile Center) has ~3,000 tons of thorium in inventory with typical concentrations of several

25
milligrams of ***Ra per ton of thorium. About 90% of this is considered excess. This implies a total
inventory of several grams of ***Ra with the potential to produce several grams of **°Th, a fraction of that
available from *°U. The primary issue with this production route is the availability of ***Ra.

While the quantity of ***Ra is relatively small, it may be possible to recover this material. The
thortum 1s in the form of soluble thorium nitrate, and the federal government is considering converting
this material to an oxide for storage or disposal. The traditional conversion process is to dissolve the
thorium nitrate in water, precipitate the thorium with oxalate, and calcine the thorium oxalate to insoluble
thorium oxide. In that process, much of the radium dissolves in solution. This may allow the low-cost

recovery of the **Ra by ion exchange or selective precipitation.
4.1.3 Availability of *’Bi
4.1.3.1 Demand

The demand for 2"*Bi for research purposes is growing rapidly. Phase I clinical trials are currently
underway at the Memorial Sloan-Kettering Cancer Center in New York City for the treatment of acute
myelogenious leukemia. The combined current production capacities at ORNL and ITU are being used
to meet this need. A large number of pre-clinical trials are ongoing and some are expected to go into
Phase I clinical trials by the end of 1999. Pre-clinical trials are underway in several locations: (1)
Memorial Sloan-Kettering Cancer Center, New York (prostate cancer); (2) National Institute of Health;
(3) University of Washington; (4) INSERM, France (multiple myeloma); (5) University of Heidelberg,
Germany (non-Hodgkins lymphoma); (6) Clinic Hasselt, University of Gent, Belgium (non-Hodgkins
lymphoma); (7) University of Gottingen, Germany (colon cancer); (8) Kantonsspital Basel, Switzerland
(low-grade glioma); and (9) Universitatsklinik, Miinchen, Germany (stomach cancer).

The demand for 2"*Bi for research exceeds the current supply. The only production-scale process
currently is extraction of *’Ac from **°Th that, in turn, was obtained from ***U. This process is being
used at ORNL and ITU. Additional **Th is being extracted from ***U to meet the demand. All research
needs could be supplied from the **Th in the *°U.

The total *°U inventory contains about 40 g of **Th, sufficient such as to produce enough *°Bi on a
continuing basis for treatment of 100,000 patients/year. A typical patient uses about 5 mCi of **Ac.
However, there are significant uncertainties. Clinical trials have not defined the preferred doses.
Furthermore, the amount of *’Ac delivered to the hospital per gram of *’Th depends upon the efficiency
of multiple chemical separation steps, packaging times, transport times, and other factors. This 1s also
sufficient to treat one major type of cancer. If *'’Bi becomes a preferred option for treatment of cancers,
the U.S. demand for *’Bi would be significantly larger than could be supplied from existing stocks of

23U, The world demand for *"*Bi would be an order of magnitude larger than the potential U.S. demand.

26
The current shortages and potential future markets are resulting in development and construction of

prototype facilities to produce *’Bi by alternative methods.
4.1.3.2 Economics

There are multiple methods to produce *"*Bi. Alternative production techniques for ?>Bi have been

identified and investigated. However, there are several economic considerations.

« Cost. No comparative estimates of the relative production costs of *’Bi from different sources
currently exists. An economic study is underway to develop an understanding of the relative
production costs of some production methods (Ehst 1999). Factors such as the efficiency of
chemical separations can strongly impact total economics because different production routes
require different chemical separations. If ?“°Bi is produced from *°U, security costs associated
with processing weapons-usable *°U are a significant fraction of total production costs. If the

only use ever to be made of >°U is for medical applications, strong incentives may exist to
isotopically dilute the 2°U with 2*U to convert the **°U to non-weapons-usable ***U.

»  Availability. There may be low-cost sources of ?’Bi; however, if the supply is limited alternative
production techniques are required. For large-scale use, the resource base for large-scale
production must be understood.

» Competition. If there is only one low-cost production route and if it is a proprietary technology,
prices may be high. In such cases, there may be an interest in maintaining the next lowest-cost
production option to produce *°Bi to limit societal costs.

»  Reliability. The production of 2°Bi from ***Th is substantially simpler than production from
other routes. Only a simple chemical separation is required. Consequently, there are reliability
advantages in producing *Bi from **Th. The **°Th can be produced from ***U or **°Ra.

4.1.4 Assessment and Conclusions

There are major uncertainties associated with the demand for ?°Bi. Uranium-233 is the current
source of **Th, which, in turn, is the source of ?"’Bi, but there are alternatives for production of *°Bi. If
2Bi becomes the preferred treatment for one or two cancers, the 2°U inventory may be able to supply
the U.S. needs for “Bi. If ?“Bi becomes the preferred treatment for several cancers worldwide, demand
will exceed supply, and other production techniques will be required. The potential demand has resulted
in several groups developing alternative production techniques and preparing plans for larger-scale
production if clinical trials on *°Bi are highly successful. Economics will ultimately determine the
preferred production method or methods.

All the U inventory, except the LWBR ***U, can be a near-term source of **Th. The CEUSP
inventory is the largest single source of *°Th, but is the most expensive source because of the low
concentration of 2°U (hence ***Th) in the uranium. If the CEUSP material is being processed for other
purposes, *°Th recovery may be desirable; however, the high cost of handling and processing this

material limits its long-term value for this application.

27
If medical applications were the only use of *°U, consideration should be given to converting it to
non-weapons-usable 2°U to reduce security costs.

Studies are underway to understand the relative production costs of some *"’Bi production routes.
These studies may provide definitive answers whether *°U is needed just in the short-term or both the
short- and long-term as a source of ?’Bi. Such studies may also determine whether the CEUSP material
1s worth saving or whether processing costs make it uneconomical under any scenario. These studies
should be expanded to include all major production options and be completed at the earliest date and

receive peer review.

4.2 REACTORS FOR DEEP-SPACE AND OTHER SPECIAL MISSIONS

Reactors fueled with 2**U can be designed with a smaller mass than either **U or #’Pu reactors, thus
there has been an interest in using *°U as a nuclear reactor fuel for deep-space missions, for which a
premium is placed on minimizing mass. For this application, only high-quality **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.
Table 4.3 summarizes this use of *°U.

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

depends upon the power, energy, and safety requirements.

» Power. For power production levels up to many kilowatts, the minimum-mass nuclear power
source is a radioisotope generator (Fig. 4.4). 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 *°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 critical masses; however, the mass of a *°U reactor, including the nuclear
moderator and other required components, is less than that for a plutonium reactor. Furthermore,
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
233U 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. To obtain the heat transfer, the reactor fuel
assemblies 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.

28
Table 4.3. Summary: reactors for deep-space and other special missions

 

Application Low-power, minimum mass reactor for deep space and other missions.
Acceptable ***U feeds High-isotopic-quality

Isotopic purity requirements  High-1sotopic-quality

Demand Limited demand unless special defense applications

The market for ***U small reactors is between ***Pu thermoelectric generators and small >°U
reactors. The unanswered question 1s whether the need for reactors in this power range 1s
sufficient such as to warrant the high development costs of 2*U reactors and the handling
difficulties of such reactors. Uranium-233 1s an alpha emitter (similar to plutonium) with the
associated safety requirements. Uranium-235 reactors avoid these complications.

Use implications Consumptive use

Technical description Uranium-233 reactors provide the minimum mass reactors over a small range of power
demands. There may be an application for **U reactors for deep-space and other special-
purpose missions for which there are very large weight penalties.

Assessment There is a potential demand, but there are major limitations including the added safety
requirements of >°U compared to **°U.

 

» FEnergy. The total mission energy requirements also impact the choice of fuel for a space reactor.
In missions with large total energy requirements, significant quantities of fissile materials must
be in the reactor to provide the energy for a long-term mission. Under such circumstances,
reactor mass 1s not determined by the choice of fissile material. HEU becomes the preferred
material.

» Safety. Uranium-233 is an alpha emitter like plutonium. Consequently, complex safety measures
are required to ensure safety in the event of a rocket launch failure from the surface of the earth
to earth orbit. This has an associated weight penalty until earth orbit 1s reached at which point
some protective devices can be ejected to minimize weight on probes going beyond earth. The
hazards of U are sufficiently low such that these additional safety measures are not required.
There is a trade-off between the weight penalties for safety features associated with 2°U vs the
weight penalty associated with 2°U.

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 high-isotopic-quality 2*U would be used. Low-isotopic-quality **U
1s unacceptable because of the weight penalty from other uranium 1sotopes. Intermediate isotopic quality
23U is highly undesirable because the radiation levels associated with this material imply the need for
remote placement and operations during preparations to launch the reactor into orbit. The weight

penalties would prohibit shielding the reactor.

29
ORNL DWG 98C-7620R

A HEU Reactor

e Reactor Mass Not Controlled by
Critical Mass
- High Burnup Fuel
- High Power Output

¢ Minimal Launch Risk

- Low-Alpha Emitter

  
   
 

Power Level

233U Reactor
e Minimum Critical E\flasfi

 

238py Heat Source
(No Minimum Critical Mass)

 

 

Mission Duration (Time)

Fig. 4.4. Minimum mass space nuclear power sources for different power levels and
mission duration.

30
4.3 ANALYTICAL MEASUREMENTS

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 2°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. Table 4.4 summarizes the characteristics of this

application of 2°U.

4.4 NUCLEAR WEAPONS RESEARCH

Because **U is fissile, it has the potential to be used in nuclear weapons (Table 4.5). There are three
potential uses:
*  Tracer. Uranium-233 may be used as a tracer to measure residual uranium after a nuclear

weapons test for any weapon with any type of uranium. This is basically the same use as that
described in Sect 4.3.

«  Weapons physics. Nuclear weapons can be made of plutonium, HEU, and **U. Test devices
may be made of 2°U to better understand how a nuclear weapon works or to calibrate computer
simulations of nuclear weapons tests. It is similar to standard test procedures used in the
development of many industrial products. For example, when developing a new paint, many
potential paints will be formulated and tested. This will include paint formulations using
ingredients that are clearly too expensive for a commercial paint. However, testing with many
paint variations develops an understanding that allows formulation of high-performance, low-
cost paints. The same logic applies to all engineered products—including nuclear weapons.

*  Weapons. Uranium-233 can be used in nuclear weapons. The current inventory would be
insufficient such as to build a large nuclear weapons arsenal. However, there is the policy option
of reserving sufficient 2°U such as to allow the development of 2*U weapons while production
of large quantities of ***U was initiated.

For all applications, only high-isotopic-quality **U would be used. For the first application, only
small quantities of **U would be required. For the other two applications, larger quantities are required.
The TAEA defines 8 kg of ***U as a significant quantity. A significant quantity is that amount recognized
by treaty as sufficient to build a nuclear weapon. For a physics test, somewhat >8 kg would probably be
required given fabrication losses. If several physics tests were desired, some multiple of 8 kg would be
needed. If 2°U weapons were to be developed, some multiple of 8 kg would be required.

Any of the previous applications would require a national decision to resume nuclear weapons
testing. Setting aside ***U for defense purposes is a contingency option in the event of major changes in

international relations.

31
Table 4.4. Summary: analytical measurements

 

Application

Acceptable ***U feeds
Isotopic purity requirements
Demand

Use implications

Technical description

Safeguards, analytical measurements.
High-isotopic quality **U

Best in the inventory

Less than 100 g/year.

Consumptive use

There are many methodologies. In the most common application, a known quantity of 2*U is
added to an unknown uranium sample. A mass spectrometer determines the ratio of different
uranium isotopes to >’U. The concentration of the different uranium isotopes is then
calculated.

 

 

Assessment This is a very important use of 2*U; however, only small quantities of 2**U are required—a
few kilograms.
Table 4.5. Summary: nuclear weapons
Applications Diagnostics for nuclear weapons tests. Material for nuclear weapons physics tests.
Alternative fissile material for nuclear weapons
Acceptable **U feeds High isotopic quality
Isotopic purity requirements ~ High 1sotopic quality

Demand

Use implications

Technical description

Assessment

There may or may not be a future demand. Uranium-233 needed only if the United States
decides to resume nuclear weapons testing and has an interest in the development of weapons
based on *’U. Potential future demand is <100 kg.

Consumptive use of ***U

The #*U is used as a diagnostic test. It could be a replacement material for plutonium or
HEU in nuclear weapons.

Minor use of >*U

 

32
4.5 REACTOR FUEL CYCLE RESEARCH

Uranium-233 is required if the United States wants to maintain the option for large-scale research

and development (R&D) on uranium-thorium nuclear power fuel cycles (Table 4.6).

Table 4.6. Summary: reactor fuel cycle

 

Application

Acceptable ***U feeds
Isotopic purity requirements

Demand

Use implications
Technical description

Assessment

Development of thortum fuel cycles including proliferation-resistant nuclear, accelerator, and
fission—fusion power reactor fuel cycles.

High-isotopic quality ***U and intermediate-quality ***U (lesser value).
Qualified materials for R&D.

Potentially large demand that could require use of the entire inventory of high-quality and
intermediate-quality ***U.

Consumptive and nonconsumptive use of *°U.
The ***U would be used for research: critical pile tests, test fuel assemblies, etc.

This is potentially the major use of **U. There is a single policy issue: Do we wish to
maintain the capability to conduct development on thorium fuel cycles including
proliferation-resistant fuel cycles? If the answer is “yes,” all high-quality 2**U should be kept.
It is highly desirable to keep intermediate-quality >**U. There is little value in the low-
isotopic-quality 2*°U.

 

4.5.1 Application

The major historical application for *°U has been for research into new nuclear power reactors and

associated fuel cycles that produce ***U from thorium. This is also a potential future application. There

are five incentives for considering a **U-thorium fuel cycle.

*  Nonproliferation. Uranium-233-thorium fuel cycles have significantly lower risks of diversion
of weapons-usable material than do conventional uranium-plutonium fuel cycles (Herring 1998).

* Resources. The global resources of thorium are about four times greater than those of uranium.
If uranium becomes scarce, thorium is a more abundant fertile material to use in reactors to breed

nuclear fuels.

o [Fuel efficiency. In thermal-neutron reactors, such as LWRs and high-temperature gas cooled
reactors (HTGRs), thorium fuel cycles breed more fissile material (**°’U) than do reactors fueled
with LEU (Ronen 1990). This reduces the demand for natural uranium per unit of energy
produced in any fuel cycle where the SNF is reprocessed and the *°U is recycled back to the
power reactors. In high-burnup, once-through fuel cycles, thorium fuels reduce the consumption

33
of natural uranium. LWRs are today the dominant type of nuclear power reactor. HTGRs are an
advanced type of power reactor.

» Waste. The performance of a thorium oxide (ThO,)-based SNF is expected to be orders of
magnitude better than uranium dioxide (UO,) LWR SNF in geological repositories that have
oxidizing conditions—such as the proposed Yucca Mountain repository. In an oxidizing
geological environment, ThO, 1s chemically inert, whereas uranium dioxide can be oxidized to
higher uranium oxides with faster release of radionuclides from the SNF. Such fuel cycles also
produce significantly smaller quantities of long-lived actinides such as plutonium, americium,
and curium.

*  Nuclear fuel performance. Thorium fuels have somewhat better thermal and mechanical
performance than do uranium fuels because of the advantageous physical properties (thermal
conductivity, fission gas retention, melting point, etc.) of thorium oxide compared to uranium
dioxide (Herring 1998).

4.5.2 Nonproliferation Fuel Cycles

The most distinctive characteristic of thorium fuel cycles with 2°U vs uranium fuel cycles is the
potential for development of more proliferation-resistant fuel cycles. These are fuel cycles for which it 1s
very difficult to recover fissile material for use in nuclear weapons. Thorium—>**U proliferation-resistant
fuel cycles (Sege, Strauch, Omberg, and Spiewak 1979; Sehgal, Naser, Lin, and Loewenstein 1977,
Herring and MacDonald 1998) have been partly developed for both once-through fuel cycles and fuel
cycles involving reprocessing of SNF. There are four characteristics that allow the development of

nonproliferation fuel cycles.
4.5.2.1 Isotopic Dilution

Uranium-233, like >**U, can be isotopically diluted with 2*U to convert the ***U into nonweapons-
usable material. In contrast, isotopic dilution can not be used to convert plutonium into a non-weapons-
usable material. Uranium-233 is made from the neutron bombardment of 22Th. If the thorium is mixed
with some *U when it is bombarded with neutrons, the *°U that is created will be isotopically mixed
with the #*U as it is generated to produce non-weapons-usable **U. This nonproliferation advantage
applies to thorium—>>*U reprocessing and once-through fuel cycles.

Isotopic dilution also provides long-term non-proliferation advantages. The barrier to proliferation
exists forever. In contrast, the primary barrier to recovery of plutonium from SNF is the initially high
radiation levels. SNF is highly radioactive; however, over time the radiation level decreases, and it
becomes progressively easier to recover plutonium from SNF. Furthermore, the quality of the plutonium

in the SNF improves with time. Plutonium in SNF contains many isotopes. The preferred material for

34
nuclear weapons is ?°Pu. It is also the long-lived isotope of plutonium. As SNF ages, the less desirable
plutonium isotopes in terms of weapons use decay, thus resulting in a better plutonium for weapons.

The addition of 2*U for isotopic dilution of 2*U does imply that some plutonium is created in the
nuclear fuel during irradiation. However, the quantities of plutonium produced may be <20% of those of
comparable Z°U-plutonium fuel cycles because most of the #*U has been replaced with thorium. In
many of these thorium—>**U fuel cycles, the isotopic composition of the plutonium is significantly less

desirable for use in nuclear weapons than is plutonium produced by typical *U—plutonium fuel cycles.
4.5.2.2 Radiation

Except when using special production techniques, significant quantities of ***U are produced as a by-
product of *U production. One of the decay products of ?U is *®*T1 with its 2.6-MeV gamma ray. This
implies that the 2°U will have significant radiation levels. If such materials are used for weapons, this in
turn implies the following: remote fabrication of nuclear weapons, RH of weapons, and radiation
damage to sensitive electronic components in nuclear weapons. Furthermore, it is difficult to hide a
weapon that emits high-energy gamma rays. This characteristic complicates security because the
radiation field makes it easier for an outsider to determine exactly where a nuclear weapon is being
stored.

Some perspectives on potential radiation levels of 2°U with high concentrations of ***U can be
obtained by examining the radiation levels of fresh (unirradiated) fuel assemblies with the *?U in secular
equilibrium with its decay products (Fig. 4.5). This occurs ~10 years after any separation that removes
decay products. For a fast reactor fuel assembly, the radiation dose at 1 ft is 1,257 R/h. The radiation
dose at 1 mis 397 R/h. This dose is from **U made in a fast reactor with SNF burnup of 100,000
MWd/t.

In terms of safeguard requirements, the IAEA recognizes that the dangers of diversion of SNF by a
nation or subnational group is significantly less than that for pure, weapons-usable fissile material.
Because different materials have different radiation levels associated with them, the IAEA has defined
SNF as fuel with a radiation level >100 R/h at a distance of 1 m (IAEA August 1993). The S&S
requirements for SNF are much less than those for separated fissile materials. It 1s noted that slightly
aged U fresh fuel with a high #?U content has a higher radiation level than does the IAEA definition of
SNF. This nonproliferation advantage applies to thorium—>**U reprocessing and once-through fuel

cycles.

35
 

 

ORNL DWG 98C-7619R3

 

 

 

 

Pressurized Water Fast Breeder Reactor
Reactor (PWR) Fuel (FBR) Fuel
2331) from PWR
e Irradiated 2%2Th IAEA Definition
¢ 50,000 MWd/i of SNF Radiation
® 1436 ppm ©2U >100R/hat1m
209 R/h
66 R/h l
\ ; 11t 1
< 1m m >
95 R/h 397 R/h
298 R/h 1257 R/h

2334 from PWR

e Irradiated %°Th
e 100,000 MWd/t

e 2052 ppm 22y

Fresh Fuel Assembly
e 25 kg 233U As Oxide
® 475 kg 2%2Th As Oxide

 

 

 

 

233U from FBR

e irradiated %*?Th
e 100,000 MWd/t

® 1727 ppm 232

Fresh Fuel Assembly
¢ 125 kg 233U As Oxide
® 375 kg 2°2Th As Oxide

Fig. 4.5. Maximum radiation levels of fresh fuel with >*U.
4.5.2.3 Off-Specification Plutonium

LWRs and HTGRs with thorium fuel cycles produce limited quantities of very-poor-quality
plutonium. Such plutonium would be difficult to use for construction of weapons. A recent study
(Herring 1998) at INEEL examined the potential for this plutonium to be used in nuclear weapons and
compared the plutonium from once-through thorium fuel cycles to plutonium generated in other fuel
cycles for LWRs.

4.5.2.3.1 Quantity

The amount of plutonium produced in a thorium fuel cycle is significantly less than that produced in
a once-through LWR fuel cycle using LEU fuel. Thorium may replace up to 80% of the uranium in such

a reactor core, thus the production of plutonium may be reduced by up to 80%.
4.5.2.3.2 Quality

The plutonium from a once-through thorium fuel cycle has unusual isotopics that make construction
of nuclear weapons very difficult. Weapons designers prefer 2*°Pu for nuclear weapons. Other

plutonium isotopes cause major problems.

*  Plutonium-238. This plutonium 1sotope produces large quantities of heat and is used in deep-
space power sources (Sect. 4.2). The concentration of >**Pu in plutonium from a once-through
thorium fuel cycle is up to 40 times that in WGP. With these **Pu concentrations, a weapon
would likely require an active cooling system to prevent the plutonium from melting or the
explosives from degrading and, thus, destroying the weapon’s ability to function.

Plutonium-241 (**'Pu). This plutonium isotope spontaneously emits large numbers of neutrons.
High, spontaneous neutron production drastically limits the probable yield of a crude nuclear
weapon. The radiation dose associated with **'Pu implies (1) significantly higher radiation doses
to security forces with the potential need for radiation shielding around the weapon during
storage and (2) easier detectability at a distance using radiation detection equipment. The **'Pu
concentration from typical thorium fuel cycles is 15 times that in WGP.

This nonproliferation advantage applies to thorium—> U recycle and once-through fuel cycles.
4.5.2.4 Once-Through Fuel Cycles

The United States, as a policy, discourages the reprocessing of SNF for recovery and recycle of
fissile materials into fresh fuel. It advocates the use of once-through fuel cycles where the SNF 1s
directly disposed of. The economics of once-through fuel cycles are improved with high-burnup fuels
that produce large quantities of energy per ton of fuel. The highest-burnup, once-through fuel cycles for
LWRs and HTGRs would use thorium fuel cycles. Successful development of such fuel cycles would

reduce economic incentives to process SNF with recovery and recycling of fissile material in fresh fuel.

37
4.5.3 Current Nuclear-Power Thorium-Fuel-Cycle Developments
4.5.3.1 Countries with Active Programs

Several countries have small efforts underway to examine thorium fuel cycles. These include the
United States, France, Japan, Russia, Canada, Brazil, and India. The incentives in examining thorium
fuel cycles are different for different countries.

*  Resources. India has limited uranium resources but the world’s largest thorium resources.

Consequently, India has historically had a long-term interest in thorium fuel cycles. The Indian
program includes (1) production and separation of 2°U and (2) operation of a small research

reactor fueled with ***U. In a similar way, Brazil’s interest is a consequence of large domestic
reserves of thorium compared to uranium.

*  Reactor technology. Canada developed Canadian Deuterium Uranium (CANDU) power reactors
and has exported these reactors to several countries. There are several unique advantages for
using a fuel containing thorium in this reactor. Consequently, Canada and several countries that
own CANDU reactors have investigated, and continue to investigate thorium-containing fuels.

«  Other. Many other countries have individuals or programs investigating thorium—>>*U fuel
cycles. Examples include France, Russia, Japan (Yamawaki et al. 1999), South Korea (Kim
et al. 1999; Kim, et al. 1999), Italy (Lombardi et al. 1999), and the Netherlands (Kloosterman
1999). The larger countries have broad programs that investigate all major fuel cycles including
thortum fuel cycles.

4.5.3.2 Once-Through Thorium Fuel Cycles

Historically, the research on thorium fuel cycles emphasized fuel cycles where the SNF was
reprocessed, °U was recovered, and the 2*U was recycled into power reactors. In the last several years,
an interest has developed in once-through thorium fuel cycles in LWRs, which are the predominant type
of nuclear power plant worldwide. In these fuel cycles, the fresh fuel is a mixture of LEU and thorium.

There are several institutional and economic reasons for the renewed interest.
4.5.3.2.1 Nonproliferation

With the end of the cold war, the Irag-Kuwait war, and other events, there 1s a renewed interest in
proliferation resistant fuel cycles. These include once-through fuel cycles from which it is difficult for

even a nation-state to recover weapons-usable fissile materials.

38
4.5.3.2.2 Uranium Resources

In a once-through nuclear fuel cycle, the only source of fissile material in fresh fuel is ***U derived
from natural uranium. If once-through fuel cycles are to be economically competitive into the future,
efficient use of the uranium 1s required. If less uranium is required to produce a given amount of energy,
the cost of uranium has less of an impact on the cost of nuclear energy. An LEU-thorium—>>*U fuel may
reduce total uranium consumption compared to a LEU fuel and hence fuel costs.

There are two types of once-through fuel cycles: LEU fuel cycles and LEU—-thorium fuel cycles.
The LEU is a mixture of 2°U (the fissile component) and **U (the fertile component). In the
LEU-thorium fuel cycles, thorium replaces most of the ***U in the fuel. In a reactor, some of the
neutrons from fissioning *°U are absorbed by the fertile material. If the fertile material is 2*U, it is
converted to 2°Pu. If the fertile material is >**Th, it is converted to >**U. In LWRs, ***U is a better fuel
than #°Pu; that is, the fission of a 2*U atom results in more neutrons that the fission of a *°Pu atom. The
additional excess neutrons can be absorbed into thorium to make more ***U and extend the life of the
fuel.

The previous considerations would suggest that LEU-thorium—>**U once-through fuel cycles would
be preferable. However, there is a problem. In a reactor core with thorium, more **°U must be initially

added to start the reactor. This leads to three possible outcomes.

»  Low-burnup once-through fuel cycle. 1f a fuel element has a relatively limited lifetime in the
reactor core, the added *°U to make the reactor work is thrown away with the SNF.

«  Fuel cycle with reprocessing. If the fuel is reprocessed with the recovery of the *°U and unburnt
233U, both fissile materials can be recycled into new reactor fuel. The total consumption of
uranium is reduced as the fissile materials are recycled.

»  High-burnup once-through fuel cycle. 1f a fuel element has a sufficiently long lifetime in the
reactor core, the initial 2°U is efficiently used and enough ***U is created and used so that the
total natural uranium needed to produce a set quantity of energy may be reduced compared to a
conventional LEU fuel in a once-through fuel cycle.

When most of the research on thorium fuel cycles was being done in the 1960s and 1970s, the
technology did not exist to build high-burnup, long-lived LWR fuels. Consequently, no interest existed
then 1n a once-though thorium fuel cycle that increases the amount of uranium to produce a unit of
energy. All the early research was on thorium fuel cycles that included reprocessing. In the several
decades since then, the burnup (lifetime) of commercial LWR fuel elements has doubled. Therefore, it 1s
now beginning to become possible to design once-through LEU-thorium fuels that may require less
uranium than traditional LEU fuels. That implies the possibility that a LEU—-thorium fuel that could be

more economical than conventional LEU fuels.

39
Historically, HTGRs have used uranium-thorium fuels. The basis for this is that typical HTGR fuels
have fuel lifetimes (burnups) several times that of LWR fuels. Thus, there has always been an incentive

to consider the use of thorium 1n such once-through fuel cycles.
4.5.3.2.3 Fuel Performance

The physical properties of thorium oxide that are important in reactor operations are significantly
better than those of uranium oxides. This simplifies fuel design. As fuel lifetimes increase, fuels with
better physical properties are desirable. The use of thorium oxide is one way to improve the thermal and

mechanical performance of the fuel.
4.5.3.3 U.S. Research Programs

In the United States, several research programs have been identified that are examining some type of
thorium—>**U fuel cycle or reactor. Most of the activities are associated with investigation of once-
through thorium— *U-LEU fuel cycles. This is an area where historically there has been very little
research. In addition to these programs, there are several smaller efforts (Brown et al. 1999; Beller et al.

1999).
4.5.3.3.1 Radkowsky Reactor

A multiyear effort has been underway to develop a once-through, pressurized-water-reactor (PWR),
LEU—thorium fuel that could be retrofitted into existing LWRs (Radkowsky December 1998;

Morozov 1999; Radkowsky 1999). The design was originated by A. Radkowsky, former Chief Scientist
of the U.S. Naval Nuclear Propulsion Division of DOE. It is a derivative of the LWBR concept tested at
the Shippingport Nuclear Power Station in the 1970s as part of the U.S. Navy nuclear program. The
reactor core 1s a “seed and blanket” fuel design with the seed part of the fuel assembly made of a uranium
zirconium metal alloy and the blanket part of the fuel assembly made of mixtures of uranium and thorium
oxides.

The reactor core design is somewhat different than that of the LWBR and reflects different design
goals. The LWBR program was to demonstrate nuclear fuel breeding in an LWR—production of more
fuel in the reactor than is consumed. The fuel was expected to be reprocessed to recover the #°U for
fabrication into a new reactor fuel. The Radkowsky reactor design goals are different.

*  Nonproliferation. The SNF is designed to minimize the potential for recovery of weapons-usable
fissile materials. The SNF 1s not to be reprocessed.

40
»  Waste management. The volume of the SNF compared to the traditional fuel cycle is to be
reduced, and the SNF is to be a better waste form.

» FEconomics. The reactor core 1s designed to maximize economics by minimizing the consumption

of uranium with high *°U production rates.

The program includes several partners [Radkowsky Thorium Power Corporation, Brookhaven
National Laboratory (BNL), and several Russian institutes led by the Kurchatov Institute in Moscow] and
1s partially funded by DOE's Initiative for Proliferation Prevention program. The U.S. funding is part of
a larger joint U.S.—Russian program to employ Russian nuclear weapons scientists. Much of the work 1s
being done in Russia. The program plans to insert 1 to 6 prototypical lead-test-assemblies into an

operating VVER-1000 (Russian PWR) by the year 2005.
4.5.3.3.2 National Laboratory, University, Fuel Fabricator Consortium

A consortium is investigating advanced, once-through LWR thorium fuels as part of the Nuclear
Energy Research Initiative of the Office of Nuclear Energy of DOE. Several other sources of funding are
also supporting this program with some assistance from commercial nuclear fuel fabricators. The
recently initiated program is titled “Advanced Proliferation Resistant, Lower Cost, Uranium—Thorium
Dioxide Fuels for Light Water Reactors.” The emphasis 1s to develop a low-cost nuclear fuel that
maintains the nonproliferation benefits of thorium fuel cycles (Kazimi et al. 1999; Herring and
MacDonald 1998; Herring and MacDonald 1999). The fuel design is similar to conventional LWR LEU
fuels except that the fuel is a mixture of ThO, and UQO, rather than just UO.,.

The program is based on calculations that suggest that as the fuel burnup in LWRs increases, at some
point a thorium-uranium fuel becomes economically preferable to an LEU fuel. It is not well understood
under what circumstances this occurs. The preliminary assessments indicate potentially a 13% fuel cost
advantage assuming a thortum—LEU fuel burnup of 72,000 MWd per ton. When the original research on
thorium—>**U fuel cycles was done in the late 1960s, the LWR fuel burnups were so low that once-
through thorium—>**U fuels were uneconomical. Fuel burnups have increased sufficiently such that we
may be approaching LWR fuel burnups where a thorium—>**U fuel is economically interesting.

The proposed fuel design is closer to existing fuel designs than is the Radkowsky fuel design; thus,
the technical uncertainties should be less. However, such designs may not as efficiently generate >°U in

the reactor core and, thus, may require more uranium feed per unit of energy produced.

41
The consortium includes two national laboratories [INEEL and Argonne National Laboratory
(ANL)], several universities (MIT, Purdue, University of Florida), and all the commercial PWR nuclear
fuel fabricators in the United States (ABB Combustion Engineering, Inc.; Framatome Technologies;

Siemens Power Corporation; and Westinghouse Electric Corporation).
4.5.3.3.3 ANL and Purdue University

ANL, in collaboration with Purdue University, is examining an advanced once-through thorium—>>*U
fuel. The fuel would be made of thorium oxide/uranium oxide microspheres dispersed in a zirconium
metal matrix for use in LWRs. Such a fuel may achieve longer lifetimes with higher burnups by
operating at lower temperatures (metal fuels have higher thermal conductivities that result in lower fuel
temperatures). This effort 1s being funded by the Nuclear Energy Research Initiative of the Office of
Nuclear Energy of DOE. The recently initiated program is titled “Fuel for a Once-Through Cycle
(Th,U)O, in a Metal Matrix.” The principal investigator is S. McDeavitt of ANL.

If the development effort is successful and an economic fuel could be manufactured, it could be used

in either of the proposed reactor fuel systems described previously.
4.5.3.3.4 BNL and Purdue University

BNL 1n collaboration with Purdue University (Takahashi et al. 1999) is examining a plutonium-
thorium fueled, fast—neutron-spectrum, boiling water reactor. The fuel would be made of plutonium and
thorium. The design objectives are to achieve a high-conversion of thorium to *°U and to reduce the
national accumulated inventory of plutonium while producing electricity. In this particular concept, the
system is designed to produce *’U with high concentrations of **?U and, thus, create a proliferation-

resistant fissile uranium with a very-high radiation level.
4.5.4 Accelerator and Fusion Reactor Development

There are several advanced energy production devices—accelerators and fission-fusion
machines—that may use some variant of a thorium->*U fuel cycle. The same incentives that exist in
nuclear reactors to use a thorium—>*U fuel cycle also exist in these systems. There are, however, some
differences. Because of the very-high energy neutrons generated in these systems, thorium may fission.
The systems produce fuel (***U) and energy.

The same technical issues in terms of thorium fuel cycles exist for these concepts as exist for nuclear
reactor thorium—>U fuel cycles. These includes issues such as nuclear criticality (Oda, Martinex-val,
and Perlado, December 1998). Consequently, *°U would likely be required for development purposes if

such machines were developed.

42
4.5.4.1 Accelerators

Several groups (Oda, Martinex-val, and Perlado, December 1998; Rubbia et al, 1995¢;

Fernandez et al., 1996; Van Tuyle 1998; Beller et al. 1998) are investigating spallation neutron sources to
produce electricity and destroy specific long-lived radionuclides in wastes. These devices are called
high-power energy amplifiers. The basic concept is to accelerate protons into a heavy metal target to
produce spallation neutrons. The neutrons then fission uranium, thorium, or plutonium. In addition to
producing energy, there is the objective of destroying long-lived radionuclides to minimize requirements
on the repository that disposes of the radioactive waste. There are several programs worldwide.

»  European Organization for Nuclear Research (CERN). The largest program is centered at
CERN which is located near Geneva, Switzerland. CERN has the largest accelerator in the
world. The research program involves multiple countries and many investigators. The research
effort is led by the former director of CERN and Nobel-prize-winner Carlo Rubbia. The

emphasis is on an accelerator concept to produce power while destroying long-lived radioactive
wastes.

* Los Alamos National Laboratory (LANL). This effort is examining a combined system of
(1) accelerators to destroy weapons-usable actinides (primarily plutonium) and long-lived
radionuclides in wastes and (2) LWRs using a modified thorium fuel cycle designed to minimize
proliferation risks and reduce uranium consumption.

4.5.4.2 Fusion-Fission Hybrids

Several types of fusion-fission hybrid machines to produce energy have been proposed
(Maniscalo 1975). In each concept, a fusion reactor produces high-energy neutrons. The neutrons then
fission uranium, thorium, or plutonium. The hybrid concepts reduce the technical demands on building a
fusion reactor. In most these hybrid systems, the fuel that is produced is used to fuel additional nuclear

reactors. It 1s a method to produce nuclear fuel if the price of uranium significantly increases.
4.5.5 Material Requirements

If the United States chose to develop a **’U-thorium fuel cycle, an inventory of 500 to 1,500 kg of
23U would be desired. This need would apply to fuel cycles for nuclear reactors, accelerators or fission-
fusion machines. Only high- and intermediate-isotopic quality material would be useful. The
intermediate isotopic quality material would be less valuable than the high-isotopic quality material. The
low-1sotopic-quality material would be of little value because the neutronic properties are partly those of

the isotopic impurities in the *°U.

43
The development of a fuel cycle may require significant quantities of high-isotopic-quality ***U for
nuclear criticality tests, prototype fuel assemblies, and other tests. Bettis Atomic Power Laboratory,
which did the initial development work on the LWBR, estimates that ~1200 kg ***U (Detrick 1998) is
required for a serious fuel cycle development program with meaningful tests. Some tests, such as nuclear
criticality tests, require an inventory of 2°U which is of the same order of magnitude as the inventory
required in a proposed nuclear reactor core. Thus, significant quantities of 2°U are required for
development of thorium—>**U fuel cycles.

The requirements for R&D are different than required for industrial operations. Consequently, high-
quality, weapons-usable ***U may be required for development of a >**U fuel cycle even if the goal is
development of proliferation-resistant fuel cycles. For example, nuclear criticality tests are used to
confirm the nuclear behavior of the reactor core and, thus, assure that the reactor will behave as
expected—including safety. Similar criticality tests would be required for accelerator-driven neutron
spallation power source. In fast neutron spectrum machines (fast reactors and accelerator-driven
spallation power sources), tests would be conducted using various concentrations of 2°U and ***U in a
mock-up of the reactor. In these facilities, the *°U and **U are packaged separately in small containers.
The behavior of different concentrations of 2°U to ***U in the test facility is determined by changing the
relative number of 2°U and ***U containers. In the laboratory, the two isotopes are kept separate to allow
testing over a wide variety of conditions. In the test facility, the 2*°U is weapons-usable and can easily be
separated from the 2*U by mechanically sorting the containers by type—a low-cost option. In a power
reactor, the two 1sotopes would be isotopically mixed in the fuel, could not be separated from each other,
and, thus, would be non-weapons-usable material.

For testing once-through thorium—>**U fuel cycles for LWRs, several types of tests may be required.
The following examples indicate why significant quantities of fissile materials are required to develop a
nuclear fuel.

Criticality tests. To understand the behavior of thorium—>’U fuels in LWRs (unlike the reactor
concepts discussed above), criticality tests would have to 1sotopically blend together mixtures of
23U, 29U, and 2*U and test the various mixtures. Isotopic blending for LWR criticality tests is
required because the neutron spectrum in LWRs is such that neutron resonance absorptions are
important with respect to safety. The importance of neutron resonances is reactor dependent. In
a LWR, the behavior of the fuel is significantly different for homogeneous vs inhomogeneous
mixtures. Because the isotopes can not be practically separated after a set of tests, this is a
consumptive use of **U. For representative conditions, up to 9 fuel assemblies may be used in a

critical assembly with a typical fuel assembly containing ~500 kg of uranium. With up to 5%
23U in a fuel assembly, a set of tests may consume >200 kg of 2*U.

44
»  Burnup tests. The current interest in once-through thorium->*U fuel cycles is with very-high-
burnup fuels. It would take 10 to 15 years to conduct such tests starting with fresh fuel. One
method to conduct accelerated testing is to estimate what the fuel composition would be at half
burnup and then produce a fuel with the neutronic behavior of partly burned SNF. In effect,
create a fuel with a mixture of 2°U, 2°U, and ?**U that is similar to a fuel that has been in a
reactor for several years and then irradiate the fuel. This type of accelerated testing can answer
many development questions; however, it rapidly consumes *°U.

4.5.6 Assessment

Only two types of nuclear fuel cycles create new fissile materials: **U is used to produce **°Pu, and
#2Th is used to produce *’U. Thorium—>’U fuel cycles can be designed to be proliferation resistant and
have other potential advantages over the alternative uranium->*Pu fuel cycles. However, the technology
1s somewhat more demanding. For example, for once-through fuel cycles, it is potentially viable only
with high-burnup fuels.

If the United States wants to maintain the option to conduct R&D of thorium—>**U fuel cycles with
reasonable expenditures of time and money, all high-isotopic-quality 2*U should be kept. It is highly
desirable to keep the intermediate isotopic quality 2°U. The low-isotopic-quality 2°U has minimal value.
It is noted that the total inventory of high- and intermediate-isotopic-quality ***U is only about 700 kg.
This is only slightly more than the minimum amount of ***U that is expected to be necessary for a reactor
development program.

A fundamental dichotomy exists between uranium—>"Pu and thorium—***U fuel cycles. The United
States will automatically maintain the option to develop new uranium-plutonium fuel cycles because of
the existing inventories of plutonium (tens of tons) and HEU (hundreds of tons) maintained for defense
purposes. These defense inventories of fissile materials are 1 to 2 orders-of-magnitude larger than
needed for development of new nuclear fuel cycles. In contrast, action is required to maintain the option

to develop, with reasonable expenditures in time and money, thorium—>>*U fuel cycles.

4.6 OTHER APPLICATIONS

No other major potential applications for 2*U have been identified. Prediction of future uses is

inherently uncertain. Additional uses may develop.

45
5. CONCLUSIONS

Uranium-233 is expensive both to produce and to store. Therefore, a decision i1s needed concerning
what >**U should be kept for future uses. Currently about 2 tons of ***U-containing material are in
inventory—half in SNF. There are many minor uses; however, the total ***U needed for these uses is
<100 kg. However, there are potentially two large uses.

«  Medical. Uranium-233 is the current source of ’Bi that is being investigated for use in treating
certain cancers. If *’Bi proves a useful medical isotope, the demand may ultimately exceed the
2BBj available from #*U. Alternative methods to produce 2"’Bi are being investigated but are not
fully developed. The relative production economics of different routes to produce *"*Bi are
currently being evaluated. The relative costs to produce **Bi from **U depend upon which lot
of U is processed because of the impurities in some of the #°U in inventory. Economics will

ultimately determine preferred production routes if the clinical trials demonstrate that *"*Bi
cancer therapy is effective. This is a nonconsumptive use of 2°U.

o Thorium-*U fuel cycle R&D. There are two basic nuclear fuel cycles: uranium fuel cycles that
produce plutonium and thorium fuel cycles that produce **U. Thorium fuel cycles have several
potential advantages including the potential for development of proliferation-resistant fuel cycles
and a more robust SNF waste form. If the United States wants to maintain the capability to do
significant R&D on these fuel cycles, all the high-isotopic-quality, and preferably all the
intermediate-isotopic-quality, >*U should be retained. This would include about half of the
separated inventory of ***U-containing materials. This is a consumptive use of *°U.

Table 5.1 summarizes the uses and the categories of 2°U that may be used for different applications.
For the low-isotopic-quality 2°U (about half the separated inventory and most of the SNF), the only
potential use 1s for medical applications. The cost of processing this material will be significantly higher
than that for other U in inventory. It is recommended that an economic study be undertaken to
determine whether this material would be economically competitive for producing medical isotopes

under any reasonable set of conditions. This study may provide a definitive basis concerning whether to

keep low-isotopic quality 2°U.

47
Table 5.1. Uranium-233 uses and applicable ***U categories

 

 

 

Isotopic quality
Use
High Intermediate Low
Medical Yes” Yes Maybe
(cancer treatment)
Space Yes No No
(deep-space reactor)
Analytical Yes No No
(safeguards etc.)
Weapons Yes No No
(test, use)
Non-proliferation fuel cycle Yes Yes No

(R&D)

 

“About half of the high-isotopic quality >**U can be immediately used as a source of the medical isotope ?°Bi. The remaining
high-quality **U is mixed with thorium which prevents practical recovery of medical isotopes. Ifit is desired to produce medical
inventories from this inventory, the thorium must be separated, the ***U aged for several years, and then initiate the recovery of

medical 1sotopes.

43
6. REFERENCES

Albright, D, F. Berkhout, and W. Walker, 1997. Plutonium and Highly Enriched Uranium 1996: World
Inventories, Capabilities, and Policies, Oxford University Press, Inc., New York, New York.

American Nuclear Society, 1981. Nuclear Criticality Control of Special Actinide Elements,
ANSI/ANS-8.15-1981, La Grange Park, Illinois.

Beller, D. E., W. C. Sailor, and F. Venneri, October 2, 1998. A4 Closed ThUOX Fuel Cycle for LWRs with
ADTT (ATW) Backend for the 21*" Century, LA-UR-98-4186, Los Alamos National Laboratory,
Los Alamos, New Mexico.

Beller, D. E., et al., September 1999. “A Closed, Proliferation-Resistant Fuel Cycle With Th-UQO,
Fueled LWRs, Th, U, and Np Recycle, and Accelerator-Driven Transmutation of Waste (ATW),” Proc.
Global °99, International Conference on Future Nuclear Systems: Nuclear Technology—Bridging the
Millennia, American Nuclear Society, La Grange Park, Illinois.

Bereolos, P. J., C. W. Forsberg, D. C. Kocher, and A. M. Krichinsky, April 1998. Strategy for Future
Use and Disposition of Uranium-233: Technical Information, ORNL/TM-13552, Lockheed Martin
Energy Research Corp., Oak Ridge National Laboratory, Oak Ridge, Tennessee.

Bereolos, P. J., L. C. Lewis, C. W. Forsberg, S. N. Storch, and A. M. Krichinsky, June 1998. Strategy
for the 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.

Brown, N. W, et al., September 1999. “The Secure, Transportable, Autonomous Reactor System,” Proc.
Global °99, International Conference on Future Nuclear Systems: Nuclear Technology—Bridging the
Millennia, American Nuclear Society, La Grange Park, Illinois.

Defense Nuclear Facilities Safety Board, March 3, 1997. Recommendation 97-1 to the Secretary of
Energy, Washington, D.C.

Detrick, C., October 16, 1998. Personal communication with L. Lewis at Idaho National Engineering
and Environmental Laboratory, Idaho Falls, Idaho, Bettis Atomic Power Laboratory,
West Milton, Pennsylvania.

Ehst, D. A., June 11, 1999. Argonne National Laboratory. Personal communication with C. W. Forsberg
of Oak Ridge National Laboratory.

Elam, K. R., C. W. Forsberg, C. M. Hopper, and R. Q. Wright, November 1997. Isotopic Dilution
Requirements for U Criticality Safety in Processing and Disposal, ORNL/TM/13524, Lockheed
Martin Energy Research Corp., Oak Ridge National Laboratory, Oak Ridge, Tennessee.

Feinendegen, L. E. and J. J. McClure, eds., May 30-31, 1996. Workshop: Alpha-Emitters for Medical
Therapy, Denver, Colorado, DOE/NE-0113, U.S. Department of Energy, Germantown, Maryland.

49
Fernandez, R., P. Mandrillon, C. Rubbia, and J. A. Rubio, February 1996. A4 Preliminary Estimate of the
Economic Impact of the Energy Amplifier, CERN/LHC/96-01 (EET), European Organization for Nuclear
Research, Geneva, Switzerland.

Forsberg, C. W., January—March 1997. “Long-Term Criticality Control in Radioactive Waste Disposal
Facilities,” Nucl. Safety, 38(1), 60—609.

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.

Forsberg, C. W., C. M. Hopper, J. L. Richter, and H. C. Vantine, March 1998. Definition of Weapons-
Usable Uranium-233, ORNL/TM-13517, Lockheed Martin Energy Research Corp., 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.

Fuger, J. J., Lothar Koch, and J. N. C. Van Geel, November 11, 1994. Method for Producing
Actinium-225 and Bismuth-213, European patent: US5355394/1L.U87684/W(09113443,
EUROATOM(LU).

Galperin, A., et al., September 1999. “A Thorium-Based Seed Blanket Fuel Assembly Concept to
Enhance PWR Proliferation Resistance,” Proc. Global 99, International Conference on Future Nuclear
Systems: Nuclear Technology—Bridging the Millennia, American Nuclear Society, La Grange Park,
[linots.

Ganguly, C. et al., October 1991. “Fabrication Experience of AI-***U and Al-Pu Plate Fuel For the
Purnima IIT and Kamini Research Reactor,” Nucl. Technol. 96(1), 72.

Geerlings, M. W_, R. van der Hout, F. M. Kaspersen, and C. Apostolides, 1993. “The Feasibility of **Ac
as a Source of Alpha-Particles in Radioimmunotherapy,” Nucl. Med. Commun., 14(2), 121.

Hall, J. C., July 22, 1998. Letter to H. R. Canter et al., Commercial Reuse of DP Excess Material Stored
at ORNL, Oak Ridge Operations Office, U.S. Department of Energy, Oak Ridge, Tennessee.

Herring, J. S. and P. E. MacDonald, November 13, 1998. Characteristics of a Mixed Thorium—Uranium
Dioxide High-Burnup Fuel, INEEL/EXT-99-00094, Idaho National Engineering and Environmental
Laboratory, Idaho Falls, .

Herring, J. S. and MacDonald, P. E., September 1999. “Mixed Thoritum-Uranium Dioxide Fuel for High
Burnup in Light Water Reactors,” Proc. Global *99, International Conference on Future Nuclear
Systems: Nuclear Technology—Bridging the Millennia, American Nuclear Society,

La Grange Park, Illinois.

International Atomic Energy Agency, August 1993. The Physical Protection of Nuclear Materials,
INFCIRC/225/Rev. 3, Vienna, Austria.

50
Kazimi, M. S.; et al., July 1999. Proliferation Resistant, Low Cost, Thoria-Urania Fuel for Light Water
Reactors, MIT-NFC-TR-018, Massachusetts Institute of Technology, Cambridge, Massachusetts.

Kim, M. H., I. Woo, and H. K. Joo, September 1999. “Advanced PWR Core Concept With Once-
Through Thorium Fuel Cycle,” Proc. Global '99, International Conference on Future Nuclear Systems:
Nuclear Technology—Bridging the Millennia, American Nuclear Society, La Grange Park, Illinois.

Kim, Y. N., J. K. Kim, and W. S. Park, September 1999. “The Neutron Spectrum Effects on Burnup
Behavior of the Thorium-Based Fuel in the Subcritical System for TRU Transmutation,” Proc. Global
99, International Conference on Future Nuclear Systems: Nuclear Technology—Bridging the
Millennia, American Nuclear Society, La Grange Park, Illinois

Kloosterman, J. L. and H. Gruppelaar, September 1999. “Multi-Recycling of Actinides in Thorium
Based Fuels,” Proc. Global ’99, International Conference on Future Nuclear Systems: Nuclear
Technology—Bridging the Millennia, American Nuclear Society, L.a Grange Park, Illinois.

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,

Koch, L., J. Fuger, and J. Van Geel, January 8, 1997a. Process for Producing Actinium-225 from
Radium-226, European Patent: EP0752710/LU88637/W091/13443(PCT).

Koch, L., J. Fuger, and J. Van Geel, January 8, 1997b. Process for Producing Actinium-225, European
Patent: EP0752709/1.LU88636.

Koch, L. June 29, 1999. Personal communication with L. R. Dole, Oak Ridge National Laboratory,
Oak Ridge, Tennessee.

Koch, L. June 30, 1999. Personal communication with L. R. Dole, Oak Ridge National Laboratory,
Oak Ridge, Tennessee.

Lantz, E. and W. Mayo, 1972. “A Small 1400EK Reactor for Brayton Space Power Systems,” Am. Nucl.
Soc. Trans., 15(1), 4.

Lombardi C., E. Padovani, M. E. Ricotti, and F Vettraino, September 1999. “Plutonia-Thoria Fuel Cycle
as Starting Solution for a Wider Thorium Use,” Proc. Global *99, International Conference on Future
Nuclear Systems: Nuclear Technology—Bridging the Millennia, American Nuclear Society,

La Grange Park, Illinois.

Ludewig, H., et al., 1989. Small Propulsion Reactor Design Based on Particle Bed Reactor Concept,
BNL-41450, Brookhaven National Laboratory, Upton, New York.

MacFarlane, D., 1963. A 200-watt Conduction-Cooled Reactor Power Supply for Space Application,
ANL-6694, Argonne National Laboratory, Argonne, Illinois.

Mangeno, J. J. and C. W. Burrows, March 1995. Occupational Radiation Exposure from Naval

Reactors’ Department of Energy Facilities, NT-95-3, U.S. Department of Energy, Naval Nuclear
Propulsion Program, Office of Naval Reactors, Washington, D.C.

51
Maniscalco, J., January 1976. “Fusion-Fission Hybrid Concepts for Laser-Induced Fusion,” Nucl.
Technol. 28, 98.

Mirzadeh, S., 1998. “Generator-Produced Alpha-Emitters,” Appl. Radiat. Isot. 49 (4), 345.

Morozov, A. G., A. Galperin, and M. Todosow, January 1999. “A Thorium-Based Fuel Cycle for
VVERs & PWRs—a Nonproliferation Solution to Renew Nuclear Power,” Nuclear Engineering
International, 44 (534), pg. 13.

Oda, A., J. M. Martinez-Val, and J. M. Perlado, December 1998. “Criticality Studies on Molten Lead
Energy Amplifiers,” Nucl. Technol. 124, 201.

Orth, D. A, April 1979. “Savannah River Plant Thorium Processing Experience.” Nucl. Technol. 43, 63.

Radkwosky, A., January 1999. “Using Thorium in a Commercial Nuclear Fuel Cycle,” How to Do It,
Nuclear Engineering International, 44(534), 14.

Radkowsky, A. and A. Galperin, December 1998. “The Nonproliferative Light Water Thorium Reactor:
A New Approach to Light Water Reactor Core Technology,” Nucl. Technol. 124,215.

Ronen, Y., ed., 1990, High Converting Water Reactors, CRC Press, Inc. Boca Raton, Florida.

Rubbia, C. et al., 1995. Conceptual Design of a Fast Neutron Operated High Power Energy Amplifier,
CERN/AT/95-44(ET), Geneva, Switzerland.

Schwartz, S. [, et. al., 1998. Atomic Audit: The Costs and Consequences of U.S. Nuclear Weapons
Since 1940, Brookings Institution Press, Washington, D.C.

Sege, C. A., S. Strauch, R. P. Omberg, and I. Spiewak, February 1979. “The Denatured Thorium
Cycle—An Overview,” Nucl. Technol., 42, 144.

Sehgel, B. R., J. A. Naser, C. Lin, and W. Loewenstein, October 1977. “Thorium-Based Fuels in Fast
Breeder Reactors,” Nucl. Technol. 35, 635.

Sherwood, D. R. and R. J. Serne, July 1993. Tuailings Treatment Techniques for Uranium Mill Waste: A
Review of Existing Information, NUREG/CR-2938, PNL-4453, U.S. Nuclear Regulatory Commission,
Washington, D.C.

Takahashi, H., U. S. Rohatgi, and T. Downar, 1999. A Proliferation Resistant Hexagonal Tight Lattice
BWR Fuel Core Design For Increased Burnup and Reduced Fuel Storage Requirements, DAT-NERI-4,
Brookhaven National Laboratory, Upton, New York.

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, 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.

52
U.S. Department of Energy, December 1998. Isotope Programs: Product & Service Catalog,
Washington, D.C.

Van Geel, J. N. C., J. Fuger, and L. Koch, February 23, 1991. Verfahreb zzur Erzeugung von
Aktinium-225 und Wismut-213, European Patent EP0443479.

Van Tuyle, G. J., June 1998. “Nuclear Applications of Accelerator-Driven Spallation Targets,” Nuc!.
Techno., 122 (3), 330.

Yamawaki, M., et al., September 1999. “Development of U-Th-Zr Alloy Hydrides As Alternative
Thorium-Base Fuel and MA Burning Target Fuel,” Proc. Global *99, International Conference on
Future Nuclear Systems: Nuclear Technology—Bridging the Millennia, American Nuclear Society,
La Grange Park, Illinois.

53
Appendix A

RADIATION LEVELS FROM **U

A-1
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. The material

may be either contact-handled (CH) or RH depending upon the radiation level. The historical dividing

line between CH and RH is 200 mrem/h at the surface of a container. The radiation levels from ***U, as

discussed in Sect. 2, depend upon (1) the impurities in *°U, particularly **U and its decay products; (2)

the age of the 2°U since the gamma-emitting decay products have been removed from the 2°U and **U

impurity; and (3) other materials with which the ***U is mixed. 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 (to minimize numerical instabilities in radiation calculations). The

following assumptions were used.

Uranium-233 impurity level. The 2*U was assumed to have 100 ppm #?U. The *°U inventory
has materials with ***U concentrations from a few parts per million to somewhat >100 ppm.

Uranium-232 age. The primary radiation from ***U in storage is from the **U decay product,
298T1. This decay product emits a 2.6-MeV gamma ray. If the uranium has been purified, the
29871 builds up over time and then decays as the 22U decays. The time of maximum radiation
levels 1s 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 1n., and the wall thickness is
1/16-1n. carbon steel. No shielding was incorporated into the drum.

Uranium chemical form. The uranium is assumed to be U0, 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 2°U with

7.407 parts DU containing 0.2 wt % ***U), the external radiation doses calculated at a distance of 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 ***U with 188 parts

DU containing 0.2 wt % *°U), the external radiation doses calculated at a distance of 1 ¢m 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).

A-3
—_
W — o0

14-18.
19.
20.
21.

22-23.
24.
25.

47.

48.

49.

50.

51.

52.

X© NN R WD =

ORNL/TM-6952

INTERNAL DISTRIBUTION
C. W. Alexander 26.  S.B. Ludwig
J. M. Begovich 27. L.E. McNeese
P. J. Bereolos 28-29.  G. E. Michaels
L. F. Blankner 30. S. Mirzadeh
L. W. Boyd 31.  H.J. Monroe
H. E. Clark 32.  D. L. Moses
E. D. Collins 33.  R.J. Moses
S. O. Cox 34.  B. D. Patton
A. G. Croff 35.  D. W. Ramey
E. G. Cumesty 36. D.E. Reichle
G. D. Del Cul 37. K.D. Rowley
L. R. Dole 38. J. E. Rushton
K. R. Elam 39. A.R. Sadlowe
C. W. Forsberg 40. B. W. Starnes
S. R. Greene 41.  S. N. Storch
M. J. Haire 42. J.R. Trabalka
W. H. Hermes 43. L. O. Wilkerson
R. F. Holdaway 44, K. A. Williams
C. M. Hopper 45.  Central Research Library
A. M. Krichinsky 46. Laboratory Records (1-RC, 1-OSTI)
EXTERNAL DISTRIBUTION

Jeff Allender, Westinghouse Savannah River Corp., Savannah River Site,
Bldg. 773-41A, Aiken, South Carolina 29808.

Joe Arango, U.S. Department of Energy, S-3.1, Rm. 6H-025,
1000 Independence Ave., S.W., Washington, D.C. 20585.

Frank W. Baxter, U.S. Department of Energy, 1000 Independence Ave, S.W._,
Washington, D.C. 20585.

D. Beller, Los Alamos National Laboratory, TSA-3, MS-F607,
Los Alamos, New Mexico 87545.

D. Bennett, Los Alamos National Laboratory,
Los Alamos, New Mexico 87545.

J. D. Bilyeu, Westinghouse Savannah River Corp., Savannah River Site,
Bldg. 703-F, Aiken, South Carolina 29808.
53.

54.

55.

56.

57.

58.

59.

60.

61.

62.

63.

64.

65.

66.

67.

68.

Scott Boeke, Westinghouse Savannah River Co., Savannah River Site,
Aiken, South Carolina 29808.

Don N. Bridges, Westinghouse Savannah River Corp., Savannah River Site,
Bldg. 703-F, Aiken, South Carolina 29808.

Mathew Bunn, Harvard University, B301, 79 J. F. Kennedy, St.,
Cambridge, Massachusetts 02138.

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.

B. A. Cook, U.S. Department of Energy, MS-1203, P.O. Box 1625,
Idaho Falls, Idaho 83415.

Ray Cooperstein, U.S. Department of Energy, Mail Stop DP-45,
19901 Germantown Rd., Germantown, Maryland 20874.

A. 1. Cygelman, U.S. Department of Energy, Office of Fissile Materials
Disposition, DOE/MD-3, Forrestal Bldg., 1000 Independence Ave., SW._,
Washington, D.C. 20585.

Bill Danker, U.S. Department of Energy, DOE/MD-3, Forrestal Bldg.,
1000 Independence Ave., S.W., Washington, D.C. 20585.

Luiz B. Da Silva, Lawrence Livermore National Laboratory, 7000 East Ave.,
Livermore, California 94550-9234.

J. W. Davidson, Los Alamos National Laboratory, TSA-3,
Los Alamos, New Mexico 87545.

William J. Desmond, Jr., U.S. Department of Energy, 1000 Independence
Ave., S.W., Washington, D.C. 20585.

Carl Detrick, Bettis Atomic Power Laboratory, P.O. Box 79,
West Mifflin, Pennsylvania 15122-0079.

M. J. Driscoll, Massachusetts Institute of Technology, Department of Nuclear
Engineering, 138 Albany St., Cambridge, Massachusetts 02139.

David A. Ehst, Argonne National Laboratory, 9700 South Cass Ave.,
Argonne, Illinois 60439.

Randy Erickson, Los Alamos National Laboratory, MS-F660, P.O. Box 1663,
Los Alamos, New Mexico 87545.
69.

70.

71.

72.

73.

74.

75.

76.

77.

78.

79.

80.

81.

82.

&3.

&4.

&5.

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.

Percy Fountain, U.S. Department of Energy, DOE/MD-63, Rm. 2051,
Cloverleaf Bldg., 1000 Independence Ave., S.W., Washington, D.C. 20585.

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, Rm. 2011,
Cloverleaf Bldg., 1000 Independence Ave., S.W., Washington, D.C. 20585.

Patricia Godoy-kain, Pacific Northwest National Laboratory, MS K8-46,
P.O. Box 999, Richland, Washington 99352,

Chuck Goergen, Westinghouse Savannah River Company, Bldg. 773-A,
P.O. Box 616, Aiken, South Carolina 29808.

Tom Gould, Lawrence Livermore National Laboratory, P.O. Box 808,
Livermore, California 94550.

Frank Graham, Westinghouse Savannah River Company, Bldg. 773-A,
P.O. Box 616, Aiken, South Carolina 29808.

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

Peter C. Green, U.S. Department of Energy, 1000 Independence Ave., S W._,
Washington, D.C. 20585.

Ehud Greenspan, University of California—Berkeley, Department of Nuclear
Engineering, Berkeley, California 94608.

Reginald Hall, Advanced Integrated Management Service, Suite 203B,
702 South Illinois Ave., Oak Ridge, Tennessee 37831.

R. Harmon, Lockheed Martin Idaho Technologies, Inc., P.O. Box 1625,
Idaho Falls, Idaho 83415.

Janet E. Hauber, U.S. Department of Energy, 1000 Independence Ave., S W,
Washington, D.C. 20585.

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

J. S. Herring, Idaho National Engineering and Environmental Laboratory,
MS-3860, P.O. Box 1625, Idaho Falls, Idaho 83415.
86.

87.

88.

&9.

90.

91.

92.

93.

94.

95.

96.

97.

98.

99.

100.

101.

102.

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.

Dave Huizenga, U.S. Department of Energy, Office of Environmental
Management, 1000 Independence Ave., S.W., Washington, D.C. 20585.

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

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.

Lothar Koch, Rieslingweg 8, W-7505 Weingarten, Germany,.

S. L. Krahn, Defense Nuclear Facilities Safety Board, Suite 700,
625 Indiana Ave., N.W., Washington, D.C. 20004.

M. S. Kazimi, Massachusetts Institute of Technology, Department of Nuclear
Engineering, 24-215B, Cambridge, Massachusetts 02139-4307.

Edward J. Lahoda, Westinghouse Electric Corporation, Science and
Technology Center, 1310 Beulah Rd., Pittsburg, Pennsylvania 15235-5098.

R. LaGrange, U.S. Department of Energy, DOE/EM-63, Rm. 2047,
Cloverleaf Bldg., 1000 Independence Ave., S.W., Washington, D.C. 20585.

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.

L. M. Lidsky, 215 Highland Ave., Newton, Massachusetts 02465-2511.

Betty Ahnde Lin, Princeton University, Center for Energy and Environmental
Studies, Engineering Quadrangle, Princeton, New Jersey 08544,
103.

104.

105.

106.

107.

108.

109.

110.

111.

112.

113.

114.

115.

116.

117.

118.

119.

O. W. Lowe, U.S. Department of Energy, 1000 Independence Ave., S W,
Washington, D.C. 20585.

P. E. MacDonald, Idaho National Engineering and Environmental Laboratory,
MS-3860, P.O. Box 1625, Idaho Falls, Idaho 83415.

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

Sean McDeavitt, Argonne National Laboratory, 9700 Cass Ave.,
Argonne, Illinois 60439.

John J. McClure, U.S. Department of Energy, 19901 Germantown Rd.,
Germantown, Maryland 20874-1290.

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

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

Morton 1. Michelson, U.S. Department of Energy, 1000 Independence Ave.,
S.W., Washington, D.C. 20585.

Ed Moore, Westinghouse Savannah River Company, Savannah River Site,
Bldg. 773-41A, Rm. 125, P.O. Box 616, Aiken, South Carolina 29808.

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

P. Niedzielski-Eihner, U.S. Department of Energy, DOE/FI-1, Rm. 5A-115,
Forestall Building, 1000 Independence Ave., S.W., Washington, D.C. 20585.

Dave Neiswander, Advanced Integrated Management Services, Inc.,
Suite 203B, 702 S. Illinois Ave., Oak Ridge, Tennessee 37830.

M. Newdorf, U.S. Department of Energy, Office of Fissile Materials
Disposition, 1000 Independence Ave., S.W., Washington, D.C. 20585

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

David Nulton, U.S. Department of Energy, Office of Fissile Materials
Disposition, 1000 Independence Ave., S.W., Washington, D.C. 20585.

Donald T. Oakley, 9612 Hall Rd., Potomac, Maryland 20854.

C. S. O'Dell, U.S. Department of Energy, DOE/EM-4, Rm. 2019, Cloverleaf
Bldg., 1000 Independence Ave., S.W., Washington, D.C. 20585.
120.

121.

122.

123.

124.

125.

126.

127.

128.

129.

130.

131.

132.

133.

134.

135.

136.

Mauren O'Dell, U.S. Department of Energy, DOE/EM-62, Rm. 2075,
Cloverleaf Bldg., 1000 Independence Ave., S.W., Washington, D.C. 20585.

K. L. Pilcher, Haselwood Enterprise, Inc., Suite 300A, 1009 Commerce Park,
Oak Ridge, Tennessee 37830.

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.

Theodore T. Saito, Lawrence Livermore National Laboratory, 7000 East
Ave., Livermore, California 94551-0808.

Linda Seward, Idaho National Engineering and Environmental Laboratory,
P.O. Box 1625, Idaho Falls, Idaho 83415.

Berry H. Smith, U.S. Department of Energy, DOE/EM-66,
1000 Independence Ave., S.W., Washington, D.C. 20585.

Paul H. Smith, U.S. Department of Energy, DOE/SC-14,
1000 Independence Ave., S.W., Washington, D.C. 20585.

George P. Smith, Jr., ABB Combustion Engineering Inc., Nuclear Operations,
1000 Day Hill Road, Windsor, Connecticut 06095.

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

J. Straalsund, Pacific Northwest Laboratory, MS K9-02, P.O. Box 000,
Richland, Washington 99352.

Steward W. Spetz, Framatome Technologies, 155 Mill Ridge Rd.,
Lynchburg, Virginia 24502.

B. Stevenson, U.S. Department of Energy, Office of Fissile Materials
Disposition, 1000 Independence Ave., S.W., Washington, D.C. 20585.

Hiroshi Takahashi, Brookhaven National Laboratory, Bldg. 475B,
P.O. Box 5000, Upton, New York 11973-5000.

John Thompson, U.S. Department of Energy, Office of Fissile Materials
Disposition, 1000 Independence Ave., S.W., Washington, D.C. 20585.

M. Todosow, Brookhaven National Laboratory, Bldg. 475B, 125 Upton Rd.,
P.O. Box 5000, Upton, New York 11973-5000.
137.

138.

139.

140.

141.

142.

143.

144.

145.

146.

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

Leo F. P. Van Swan, Siemens Power Corporation, 2300 Horn Rapids Rd.,
Richland, Washington 99352.

H. Vantine, Lawrence Livermore National Laboratory, 7000 East Ave.,
Livermore, California 94550.

Mike Walrath, Idaho National Engineering and Environmental Laboratory,
P.O. Box 1625, Idaho Falls, Idaho 83415.

Dennis W. Wester, Pacific Northwest National Laboratory, MS P7-52,
P.O. Box 999, Richland, Washington 99352,

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.

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