PO e
-;9114
oer 1978
Thorium Fuel Cycle
Alternatives
234'
I230
-------
EPA 520/6-78-008
(UCB-NE-3227)
THORIUM FUEL-CYCLE ALTERNATIVES
T. H. Pigford
C. S. Yang
Department of Nuclear Engineering
University of California
Berkeley, California 94720
Prepared for
U.S. Environmental Protection Agency
under Contract No. 68-01-1962
November, 1978
Project Officer
Bruce J. Mann
Office of Radiation Programs, LVF
P. 0. Box 15027
Las Vegas, Nevada 89114
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EPA REVIEW NOTICE
This report has been reviewed by the Office of Radiation Programs,
U.S. Environmental Protection Agency (EPA) and approved for publication.
Approval does not signify that the contents necessarily reflect the views and
policies of the EPA. Neither the United States nor the EPA makes any warranty,
expressed or implied, or assumes any legal liability or responsibility of any
information, apparatus, product or process disclosed, or represents that its
use would not infringe privately owned rights.
11
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FOREWORD
The Office of Radiation Programs carries out a national program designed
to evaluate the exposure of man to ionizing and nonionizing radiation, and to
promote the development of controls necessary to protect the public health and
safety and assure environmental quality.
Part of this program is devoted to an examination of existing and proposed
energy technologies with respect to radiological health impacts. In recent
years, a number of studies under government and private sponsorship have been
made to examine nuclear fuel-cycle alternatives to the uranium-oxide-fueled
light-water reactor presently used in the United States. These have been
motivated by a number of considerations, among which include a search for ways
to extend the nation's fission-fuel resources, as well as the examination of
various fuel-cycle alternatives in terms of nuclear explosives safeguards and
nuclear weapons proliferation issues.
Thorium-based fuel cycles have the potential for extending nuclear energy
resources. However, fuel cycles which utilize thorium may have features which
are significantly different from the uranium-oxide light-water-reactor fuel
cycle. As new fuel cycles are examined, major consideration must be given to
environmental and safety aspects. A first step in this analysis is the
development of descriptions of the basic features of proposed and potential
fuel cycles, which includes the identification of the various radionuclides
associated with these fuel cycles. It then becomes possible to examine
environmental control requirements in a preliminary fashion and to make
comparisons between alternative fuel cycles on the basis of these requirements.
The present report provides a basic reference document for the above
purpose. Comments on this analysis as well as any new information would be
welcomed.
W. D. Rowe
Deputy Assistant Administrator
for Radiation Programs
m
-------
ABSTRACT
Actinide material quantities and lifetime uranium ore requirements are
calculated for thorium fuel cycles in pressurized-water reactors, high-
temperature gas-cooled reactors, and pressure-tube heavy-water reactors, and
are compared with similar quantities for reference uranium-piutonium fueling
in light-water reactors and in fast breeders. Flowsheets are presented for
national-international fuel cycles for safeguard controls, including dispersed
national reactors fueled with thorium and denatured uranium. Long-term
radioactivity properties of high-level radioactive wastes are compared. Also
compared are the production of 1%C, 3H, 232U, and other activated radionuclides
from these reactors and fuel cycles.
-------
CONTENTS
page
1. Introduction 1-1
2. Actinide Reactions 2-1
3. Fuel Cycles for Light-Water Reactors 3-1
3.1 Uranium Fueling as a Reference Case 3-1
3.2 Thorium-Fueled Pressurized-Water Reactors 3-8
3.3 Resource Requirements for Pressurized-Hater
Reactors 3-11
4. Fuel Cycles for Heavy-Water Reactors 4-1
4.1 Uranium Fueling as a Reference Case 4-1
4.2 Thorium-Fueled Heavy-Water Reactors 4-6
5. High-Temperature Gas-Cooled Reactor 5-1
5.1 Reactor Characteristics 5-1
5.2 HTGR Fueled With Thorium and Denatured Uranium,
No Reprocessing 5-11
5.3 Resource Utilization by Current and Modified
HTGR Designs 5-14
6. Fuel Cycles for Fast-Breeder Reactors 6-1
6.1 The Reference PuO?-UO~ LMFBR 6-1
6.2 Fast Breeder Start-up with 235U 6-5
6.3 Summary of Resource Requirements for the Reference
LMFBR 6-5
6.4 Thorium Fuel Cycles for Fast Breeder Reactors 6-8
7. Technical Safeguards Features of Thorium Fuel Cycles and
Denatured Fuel Cycles 7-1
7.1 Safeguards in Normal Thorium Fueling 7-1
7.2 Low-Enrichment Denatured-Uranium Fuel Cycles 7-1
7.3 Denatured-Uranium-Thorium Cycles with Pressurized-
Water Reactors 7-2
7.4 Denatured-Uranium-Thorium Cycle with National PWR
and International LMFBR 7-6
7.5 National and International Fast Breeders 7-9
7.6 Denatured-Uranium-Thorium Stowaway Cycle for HTGR 7-11
7.7 Denatured Uranium-Thorium Cycles with National
Heavy-Water Reactors 7-11
7.8 Enrichment Vulnerability of Denatured-Uranium Fuel 7-14
7.9 Comparison of Denatured-Uranium Fuel Cycles 7-15
8. Radioactivity, Long-Term Toxicity, and Actinide Content
of High-Level Radioactive Wastes 8-1
8.1 Introduction 8-1
8.2 Radioactive Wastes From the Reference U-Fueled
Light-Water Reactor 8-1
-------
page
8.3 Waste Toxicities in Perspective 8-5
8.4 Effect of Pu Recycle on High-level Waste Toxicity 8-6
8.5 Toxicity of Unreprocessed Uranium Fuel 235 8-6
8.6 High-level Wastes from the PWR Fueled with U,
Th, and Recycled U 8-8
8.7 High-level Wastes from the PWR Fueled with Pu,
Th, and Recycled U 8~8
8.8 High-level Waste from the Uranium-Fueled and
Thorium-Fueled Heavy-Water CANDU Reactors 8-11
8.9 High-level Wastes from the Reference 235u-Th-Fueled
HTGR 8-14
8.10 Comparison of Actinide-Sources in High-level Wastes
From Alternate Fuel Cycles 8~18
9. Generation of 14C, H, and Other Radionuclides 9-1
9.1 Carbon-14q 9-1
9.2 Tritium (JH) 9-8
9.3 Sulfur-35, Phosphorous-33, and Chlorine-36 in
HTGK Fuel 9-14
-V1 Non-Volatile Radionuclides Activated in Fuel
F-lement Structure 9-16
9.5 232y in Uranium Recovered From Irradiated Thorium 9-18
1U. Summary and Conclusions lU-1
11. Acknowledgments 11-1
12. References 12-1
13. Nomenclature 13-1
Appendix A: Storage Time tor Thorium Recovered From
HTGR Fuel Reprocessing /\_1
Appendix B: Tables of Actinides in CANDU Fuel Cycles B-l
Appendix C: Calculational Methods c-1
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LIST OF FIGURES
2.1 Actinide chains in thorium fuel 2-2
2.2 Radioactive decay of natural thorium 2-4
2.3 Growth of beta activity and gamma dose due
to 232U in Uranium 2-5
2.4 Actinide chains in U and Pu fuel 2-7
3.1 Lifetime-average annual quantities for
uranium-fueled PWR with no fuel reprocessing 3-2
3.2 Lifetime-average annual quantities for
uranium-fueled PWR with fuel reprocessing
and uranium recycle 3-4
3.3 Lifetime-average annual quantities for
uranium-fueled PWR with self generated
Plutonium recycle 3-5
3.4 Lifetime-average annual quantities for
PWR fueled with plutonium and natural uranium 3-9
3.5 Lifetime-average annual quantities for
PWR fueled with uranium and thorium 3-10
3.6 Lifetime-average annual quantities for
PWR fueled with thorium, plutonium, and
recycled uranium 3-12
4.1 Annual quantities for natural-U-fueled
CANDU reactor 4-2
4.2 Annual quantities for slightly enriched
U-fueled CANDU reactor 4-3
4.3 Annual quantities for equilibrium U-fueled
CANDU reactor, with self generated Pu recycle 4-4
4.4 Annual quantities for equilibrium 235U-Th-
fueled CANDU reactor, with U recycle 4-7
4.5 Annual quantities for equilibrium Pu-Th-
fueled CANDU reactor, with U recycle 4-8
4.6 Cumulative Requirement of uranium for the Pu-topped,
thorium-fueled self sufficient CANDU reactor 4-10
5.1 Annual quantities for the near-equilibrium
235U-Th-fueled HTGR, with U recycle 5-2
5.2 Detailed annual mass flow sheet for the near-
equilibrium 235U-Th-fueled HTGR, with U recycle 5-3
5.3 Annual quantities for the denatured-U-Th-fueled
HTGR, with no recycle 5-12
-------
6.1 Annual quantities for LMFBR fueled with
natural or depleted uranium 6-2
7.1 Annual quantities for LWR cycle for inter-
national safeguards, national reactors
fueled with low enrichment (denatured)
uranium 7-4
7.2 Annual quantities for LWR cycle for inter-
national safeguards, national reactors fueled
with thorium and denatured uranium 7-5
7.3 Annual quantities for LWR cycle for inter-
national safeguards, national reactors fueled
with thorium and denatured uranium, inter-
national reactors fueled with thorium and
plutonium 7-7
7.4 Annual quantities for national PWR fueled with
thorium and denatured uranium, international
LMFBR produces make-up 233U 7-8
7.5 Annual quantities for national CANDU reactor
fueled with thorium and denatured uranium,
international Pu-burning PWR 7-10
7.6 Annual quantities for national CANDU reactor
fueled with thorium and denatured uranium,
international LMFBR produces make-up 233U. 7-13-
8.1 Pu radioactivity in high-level wastes from
U-fueled PWR 8-2
8.2 Actinide radioacitivty in high-level wastes
from U-fueled PWR 8-2
8.3 Ingestion toxicity of high-level wastes from
U-fueled PWR 8-3
8.4 Relative ingestion toxicity of fuel-cycle
residuals from U-fueled PWR 8-4
8.5 Ingestion toxicity of high-level wastes from
various fuel cycles 8-4
8.6 Pu radioactivity in high-level wastes from
235U-Th-fueled PWR with U recycle 8-7
8.7 Actinide radioactivity in high-level wastes
from 235U-Th-fueled PWR with U recycle 8-7
8.8 Ingestion toxicity of high-level wastes
from 235U-Th-fueled PWR with U recycle 8-9
vi n
-------
8.9 Ingestion toxicity of fuel cycle residuals
from 235U-Th-fueled PWR with U recycle 8-9
8.10 Pu radioactivity in high-level wastes from
Pu-Th-fueled PWR with U recycle 8-10
8.11 Actinide radioactivity in high-level wastes
from Pu-Th fueled PWR with U recycle 8-10
8.12 Ingestion toxicity of high-level wastes from
Pu-Th-fueled PWR with U recycle 8-10
8.13 Pu radioactivity in natural-U-fueled CANDU
reactor discharge fuel 8-12
8.14 Actinide radioactivity in natural-U-fueled
CANDU reactor discharge fuel 8-12
8.15 Ingestion toxicity of natural-U-fueled CANDU
reactor discharge fuel 8-12
8.16 Pu radioactivity in high-level wastes from
235U-Th-fueled CANDU reactor with U recycle 8-13
8.17 Actinide radioactivity in high-level, wastes
from 235U-Th-fueled CANDU reactor with U
recycle 8-13
8.18 Ingestion toxicity of high-level wastes from
235U-Th-fueled CANDU reactor with U recycle 8-13
8.19 Pu radioactivity in high-level wastes from
Pu-Th-fueled CANDU reactor with U
recycle 8-15
8.20 Actinide radioactivity in high-level wastes
from Pu-Th-fueled CANDU reactor with U and Ptr
recycle 8-15
8.21 Ingestion toxicity of high-level wastes from
Pu-Th-fueled CANDU reactor with U
recycle 8-15
8.22 Pu radioactivity in high-level wastes from
235U-Th-fueled HTGR with U recycle 8-16
8.23 Actinide radioactivity in high-level wastes
from 235U-Th-fueled HTGR with U recycle 8-16
8.24 Ingestion toxicity of high-level wastes from
235U-Th-fueled HTGR with U recycle 8-17
8.25 Ingestion toxicity of fuel-cycle residuals
from 235U-Th-fueled HTGR with U recycle 8-17
IX
-------
LIST OF TABLES
page
3.1 Actinides In The Fuel Charged To U-Fueled PWR 3-3
3.2 Actinides In The Fuel Discharged from the U-Fueled PWR 3-3
3.3 Actinides In The Fuel Charged To The PWR With
Self-Generated Pu Recycle 3-6
3.4 Actinides In The Fuel Discharged From The PWR With
Self-Generated Pu Recycle 3-7
3.5 Actinides In The Fuel Charged To The U-Th Fueled
PWR 3-13
3.6 Actinides In The Fuel Discharged From The U-Th
Fueled PWR 3-14
3.7 Actinides In The Fuel Charged To The Pu-U-Th Fueled
PWR 3-15
3.8 Actinides In The Fuel Discharged From the U-Th
Fueled PWR 3-16
3.9 30-Year Lifetime Ore Req-uirements For Pressurized
Water Reactors 3-17
4.1 30-Year Lifetime Ore Requirements For Heavy Water
Reactors 4-5
5.1 Actinides In The Fuel Charged To The U-Fueled HTGR 5-4
5.2 Actinides In Discharged U-Th First Cycle Make-Up
HTGR Fuel 5-5
5.3 Actinides In Discharged U-Th Second-Cycle HTGR Fuel 5-6
0-3-3
5.4 Actinides In Discharged "^U-Th HTGR Fuel 5-7
5.5 Actinides in Discharge Thorium Fuel 5-8
5.6 HTGR Fuel Particle Descriptions 5-10
5.7 Effect Of Reprocessing Cross-Over On The Composition
Of Recycled Uranium For The HTGR Equilibrium Fuel
Cycle 5-13
-------
5.8 Actinides In The Fuel Charged To The Denatured
HTGR 5-15
5.9 Actinides In The Fuel Discharged From The Denatured
HTGR 5-16
5.10 30-Year Lifetime Ore Requirements for High-Temperature
Gas-Cooled Reactors 5-17
5.11 Conversion Katio Improvements Possible For the HTGR
235
Fueled With U, Th, and Recycled Uranium 5-19
6.1 Fissile, Ore, and Enrichment Requirements To Start
A First-Generation Fast Breeder Reactor With Water-
Reactor Plutonium 6-4
6.2 Fissile, Ore, and Enrichment Requirements To Start
A First-Generation Fast Breeder Reactor on Enriched
Uranium 6-6
6.3 30-Year Lifetime Ore and Enrichment Requirements For
Fast-Breeder Reactors 6-7
6.4 Comparison of Pu-U and U-Th Fueling in LMFBR's 6-10
7.1 Comparison of Fuel Cycle Quantities for Denatured
Fuel Cycles 7-16
8.I Comparison of Actinide Quantities in High-Level
Wastes from Alternate Fuel Cycles 8-21
9.1 14C In Discharge Fuel 9-4
9.2 Estimates Tritium Production In The Coolant Of A
1000 Mwe Pressurized Water Reactor 9-9,
9.3 Summary of Tritium Production In Reactors 9-13
9.4 Additional Volatile Kadionuclides In HTGR Discharge
Fuel 9-15
9.5 Nonvolatile Radionuclides In Discharge Fuel From
Neutron Activation 9-17
232
9.6 Summary of Calculations of U In Recycled Uraniun
Recovered From Irradiated Thorium 9-19
B.I Actinides In The Fuel Charged To The Natural Uranigm-
11 Fueled" CANDU Reactor B-l
B.2 Actinides In The Fuel Discharged From The Natural
Uranium-Fueled CANDU Reactor
B-2
xi
-------
Pa^e
B.3 Actinides In The Fuel Charged To The 1.2% U-
Fueled CANDU Reactor B-3
B.4 Actinides In The Fuel Discharged From The 1.2%
235U-Fueled CANDU Reactor
B-4
B.5 Actinides In The Fuel Charged To The U-Fueled
CANDU With Self-Generated Pu Recycle
3-5
B.6 Actinides in The Fuel Discharged From The U-
Fueled CANDU with Self-Generated Pu Recycle
6-6
B.7 Actinides In The Fuel Charged To The U-Th-
Fueled CANDU Reactor B-7
B.8 Actinides In The Fuel Discharged hrom the U-Th-
Fueled CANDU Reactor B -8
B.9 Actinides In The Fuel Charged To The Pu-U-Th-
Fueled CANDU Reactor b-9
B.10 Actinides In The Fuel Discharged From The Pu-U-Th-
Fueled CANDU Reactor- B-10
-------
EPA 520/6-78-008
UCB-NE 3227
THORIUM FUEL-CYCLE ALTERNATIVES
T. H. Pigford
C. S. Yang
Department of Nuclear Engineering
University of California
Berkeley, California 94720
1. Introduction
The purpose of this report is to summarize features of alternative
power reactor fuel cycles utilizing thorium. This is a follow on to
an earlier study, whereby the fuel cycle material quantities and envir-
onmental effluents from the thorium-uranium HTGR fuel cycle were analy-
zed.
-------
current and near-term estimates of the costs of uranium ore and of
fuel cycle operations. In light-water and heavy-water reactors,
current and near-term costs favor uranium fueling, but future
higher costs of less concentrated uranium ores may eventually tilt
the choice towards thorium.
The current national interest in thorium fuel cycles is
directed towards:
(a) improved utilization of uranium resources
(b) use of thorium with denatured uranium as a possible means
of reducing the threat of international proliferation of nuclear
explosives.
Since these considerations may strongly influence the choice of a
particular fuel cycle, with its concomitant environmental problems,
the features of each of these thorium fuel cycles with respect to
resource utilization and proliferation are also discussed in this
respect.
This report also presents comparisons of the radioactive wastes
which result from these fuel cycles, with emphasis upon the actinide
content of high-level wastes. ' Differences in the production rate of
tritium, llfC, and other activated species present in discharge fuel-
reprocessing wastes are also considered.
The report first establishes, as a basis for comparison, the
principal fuel-cycle quantities for uranium fueling in light-water
reactors. Possible flowsheets for adapting these light-water reactors
to thorium fueling are then described and resource requirements and
radioactive waste properties are compared. Similar comparisons to
the base case of uranium fueling are made for the heavy-water reactors
and HTGR reactors. The results of these comparisons are summarized in
Section 10. Details of the computational methods used in the study are
summarized in Appendix C.
1-2
-------
2. Actlnide Reactions
Since most of the important differences in the characteristics
of fuel cycles with and without thorium result from differences in
actinide composition of the fuel, these differences can best be
followed by first examining the actinide reactions in thorium fuel.
Actinide reactions for uranium-plutonium fuel have been described
elsewhere (P2). The principal actinides involved in using thorium-
uranium fuel are shown in the actinide chain of Figure 2.1. The
most important reactions are the fission of 233U and 235U and the •
absorption of neutrons in 232Th to form 233U.
The relatively long 27.0-day half life of 233Pa, the precursor
of 233U, may affect the time that irradiated fuel must be stored prior
to reprocessing. If the discharged fuel is stored only for 150 days,
as is frequently specified for sufficient decay of 131I, some of
the 233Pa will remain during reprocessing. Protactinium is one of the
most difficult of the elements to separate from uranium, and the
high radioactivity of protactinium may contribute to the problem of
decontaminating the uranium product after it is separated from the
fission products and thorium. Also, for a short period of pre-repro-
cessing storage, 233Pa would have to be recovered or else its loss
would represent an appreciable fissile loss in the fuel cycle. Another
effect of the relatively long half life of 233Pa is the build-up of
233U in reactor fuel due to 233Pa decay after shutdown, thereby
adding to requirements for reactivity control.
Another problem of the thorium fuel cycle results from the radio-
activity of 72-yr 232U and its daughters (Bl). 232U is formed by
(n,2n) reaction with 232Th according to:
232-Th
and by
233[J n»2n> 232(j
It is also formed by the chain initiating with 235U:
236U IbX> 237u _ 237Np Q^ 236Np 236pu
Also, many thorium ores as well as thorium which is obtained as a by-
product of uranium mining contain traces of 230Th, a radionuclide in
the decay chain of 238U. Neutron absorption in 230Th also results in
the formation of 232U:
230Th n£U 23iTh S 23iPa n*JU 232Pa 232U
2-1
-------
237.
237
233
T'l^H^
232
T.H.Piyford
1977
Fig. 2.1 Actinide chains in thorium fuel
2-2
-------
Although significant alpha activity results from 232U in the
u to be recovered and recycled, more of a problem results from
the 2^U daughters. The 232U decay daughter is 1.91-yr 228Th, a
radionuclide which is also formed by the radioactive decay of 232Th.
As shown in Figure 2.2, the decay daughters of 228Th are all short-
lived, so they reach secular equilibrium with 228Th after a delay time
of only a few days. The decay of 212Bi and 208T1 are accompanied
by very energetic and penetrating gammas, so gamma shielding is re-
quired when fabricating fuel from recycled uranium containing 232U.
Although chemical reprocessing yields essentially pure uranium,
storage after separation and time elapsed in shipping to fabrication
allow the build-up of 228Th and its decay daughters. Consequently,
the gamma activity in separated uranium containing 232U increases
continuously with storage time, until it reaches a maximum at about
ten years after separation. The calculated growth in activity and
gamma dose rate for uranium metal containing 100 ppm 232U is shown in
Figure 2.3. As shown later, 232U concentrations in uranium recovered from
irradiated thorium may vary from a few hundred to a few thousand
parts per million.* Once uranium has been separated from thorium by
Thorex partitioning, there is considerable incentive to complete the
uranium purification and fuel fabrication quickly to avoid the in-
creasing radiation due to the build up of 228Th. Hydrogenous shielding
is also necessary because of the high-energy neutrons from alpha
decay in recycled uranium. The alphas from the decay of 233U, 232U,
and 228Th interact with light elements such as oxygen and carbon to
form neutrons, so the neutron activity also increases with storage
ti me.
The 228Th appearing with the separated thorium results in
appreciable radioactivity in the thorium. Consequently, as discussed
in Appendix A, it may not be practicable to recycle the recovered
thorium until it has been stored for about 3 to 16 yr, depending upon
the radioactivity of the uranium with which it is to be used in
fuel fabrication.
When 235U is used as fissile make-up in the thorium cycle, as in
the reference HTGR fuel cycle, the high burn up and uranium recycle
result in considerable production of 237Np, according to the reactions
shown in Figure 2.1. Neutron absorption in 237Np then results in a
relatively large activity of 238Pu. The plutonium activity is impor-
tant because of the problems of decontaminating uranium from plutonium
when reprocessing the uranium. Also, even though fissile plutonium
is formed by neutron absorption in the 238U accompanying the highly
Calculated quantities of 232U in various fuel cycles are summarized
in Section 9.5 of Chapter 9.
2-3
-------
3.10m 82
Fig. 2.2 Radioactive decay of natural thorium
2-4
-------
Basis; I kg of uranium
containing 100 ppm 232U
10 I02 I03
Time After Separation, days
Fig. 2.3 Growth of beta activity and gamma dose
pop
due to U in Uranium
2-5
-------
enriched 235U make-up, as shown in Figure 2.4, the high activity
of <^BPu may discourage the utilization of the fuel value of
Plutonium in the discharge fuel.
When 235U is used as fissile make-up in thorium-uranium fuel,
relatively little 239Pu, 2£»°Pu, 2lflPu, Am, and Cm are formed.
However, when plutonium is used as fissile make-up in a thorium fuel
cycle considerable quantities of americium and curium are formed.
These are the radionuclides which are the greatest contributors to
radioactivity and potential toxicity of the high-level wastes after
about 600 years of waste isolation, when most of the fission products
have decayed. The effects of actinides upon the long-term radio-
activity properties of high-level' radioactive wastes from the various
fuel cycles are considered in more detail in Chapter 8.
2-6
-------
1
3.79 1 10^
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Fig. 2.4 Actinide chains in U and Pu fuel
2-7
-------
3- Fuel Cycles for Light-Hater Reactors
3.1 Uranium Fueling as a Reference Case
To establish a reference for comparison with thorium cycles,
the familiar fuel cycle characteristics of uranium-fueled light-
water reactors are illustrated in Figures 3.1, 3.2, 3.3. These
are simplified versions of more detailed flowsheets wherein process
losses have been taken into account'. They have been derived from
cycle-by-cycle data calculated by Shapiro et al_. (SI) for a
pressurized-water reactor. At present in the United States the only
operable fuel cycle is the non-reprocessing cycle shown in Figure 3.1,
wherein the reactor discharge fuel is stored in water canals.
Prior to the administrative decision to defer fuel reprocessing,
the fuel cycle involving reprocessing with uranium recycle (Fig-
ure 3.2) with storage of the recovered plutonium could have been
operable in the U.S. upon completion of the generic licensing
decision on reprocessing with fuel recycle, originally scheduled
for mid 1977, and upon final licensing of the Barnwell plant, now
the only U.S. facility for commercial nuclear fuel processing.
Assuming a new and affirmative decision to proceed with reprocessing,
recycle of plutonium as well as uranium could not begin until a
facility to convert Pu(NOJ4 to PuOp is constructed at the Barnwell
reprocessing plant. This would require additional funding and about
four years for construction. Reprocessing at Barnwell with uranium
recycle and PU(NO,)4 storage could begin earlier, but it would be
limited to about f5 months at full throughput of 1500 Mg/yr,
because of limitations on Pu(N03). storage capacity (Cl). There-
fore, U.S. reprocessing with both uranium and plutonium recycle
does not seem possible until the early 1980's, and it may be
deferred beyond that date because of the delays which have been
imposed by the federal administration. Thus, it appears that the
commercial fuel reprocessing facilities in England and France may
continue as the only means of reprocessing power reactor fuel for
many years to come. For many years the U.S. reactors will operate
on the non-reprocessing cycle, requiring the construction of
additional and enlarged facilities for storing discharge fuel at
reactor sites as well as centralized discharge-fuel storage facilities.
Even though storage of discharge fuel does not appreciably
detract from the economic benefit of nuclear power in the United
States (P3), there will remain considerable incentive to proceed
with reprocessing. The principal motivations in the U.S. for
reprocessing are (1) to achieve the economic benefits from reprocessing
and uranium and plutonium recycle, (2) to reduce the required con-
sumption of uranium ore, (3) to reduce the required uranium-enrichment
capacity, (4) to provide plutonium when needed to start the breeder
reactors, and (5) to obtain additional commercial-scale experience
3-1
-------
oo
i
ro
3°/
28.5
Natural
Uranium
0.715% 235U
!69Mg
Fuel
Fabrication
ii
^235u
Mg
i
Conversion
and
Isotope
Separation
28.1
Mg
*^~Se|
1000 Mw
Light Water
Reactor
E= 30.4 Mw day/kg
Fuel Life = 3yr
0 = 0.342
L=0.80
parative Work
!08 Mg
Pigford-Yong, 1
Discharge
•* ruei
Storage
235
0.25% U
r. i
* 14! Mg
Fig. 3.1 Lifetime-average annual quantities for
uranium-fueled PWR with no fuel reprocessing
-------
TABLE 3.1. Actinides in the Fuel Charged to U-Fueled PWR
(1000 Ntoe, With or Without U Recycle!/)
Radionuclide
Uranium^/ 235
238
Total
a/ 30.4
kg/yr
8.43 x 10+02
2.78 x 10+04
2.81 x 10+04
CLZvr
1.81
9.09
a = 1.09 X 101
Weight %
3.00
97.00
100.00
Mw-day/kg HM, 34.2% thermal efficiency, 80% capacity
factor, near-equilibrium fuel cycl
- 234u
TABLE 3. 2
Radionuclide
Urani um^ 235
236
238
Total
Plutonium^ 239
240
241
242
Total
is not included
. Actinides in the Fuel
(1000 Mwe, U Fuel With
kg/yr
2.25 x 10+02
1.06xlO+02
2.66 x 1Q+04
2.69 x 10+04
1.40 xlO+02
5 .70 x 10+01
3.50 x 10*01
1.20 x 10+01
2.44 x 10+02
e.
Discharged from the U-Fueled
or Without U Recycle,1!/)
Ci/yr
4.82 x 10"01
6.72
8.87
a = 1.61 x 10+01
8.59 x 10+03
1.29 x 10+04
3.94 x 10+06
4.68 x 10+01
a = 2.15 X 10+°4
PWR-/
Weight %
0.83
0.39
98.78
100.00
57.39
23.05
14.43
5.13
100.00
3 = 3.94 x 10+06
- immediately after discharge
-/ 30.4 Mw-day/kg HM, 34.2% thermal efficiency, 80% capacity
factor, near-equilibrium fuel cycle.
C/ 23itn 237n *nrl 239|| a ra nnt included
239U are not
d/ 236Pu, 238Pu, and 2L|3Pu are not included
3-3
-------
Fuel
Fabrication
co
3%2MU
285 Mg
Natural
Uranium
Q7I5%235U
141 Mg
28.1
Mg
1000 Mw
Pigford-Yong.1977
Light Water
Reactor
E=30.4 Mw day/kg
Fuel Life = 3yr
n = 0.342
L = 0.80
Fuel
Reprocessing
Fission
Products.
0.912 Mg
Conversion
and
Isotope
Separation
Uranium Recycle
22.4 Mg, 0.83 %235U
$—-Separative Work
105 Mg
0.25%235U
135 Mg
To
Plutonium
Storage
71 % Fissile
0.240 Mg
Fig. 3.2 Lifetime-average quantities for uranium-fueled PWR
with fuel reprocessing and uranium recycle (E=fuel exposure,
n=overall thermal efficiency, Incapacity factor)
-------
Plutonium Recycle
Piqford-Yong. 1977
CO
I
en
Natural
Uranium
0.7l5%Kj
M6Mg
\ f
U-PuFuel
Fabrication
UFuel
Fabrication
61% Fissile
0.435 Mg
1000 Mw
a^M,S 1 iQht Water
Reactor ^
E=30.4MwdayAg
Fuel Life = 3yr Rep
^n = 0.542 *"
22.0 L=0.80
h'/in _.._...
Mg
|22°3Mg
Conversion
and
Isotope
Separation
1
a Uranium Recycle r
0.83%235U
IS.OMg
*-^^ Separative
Work
8 1 .7 Mg
Fission
Fuel Products
rocessing 0.9l2Mg
Depleted
" Uranium
0.45%235U
5.49Mg
I 0.25%235U
7105 Mg
Fig. 3.3 Lifetime-average annual quantities for uranium-fueled
PWR with self-generated plutonium recycle (E=fuel exposure, n=
overall thermal efficiency, L=capacity factor)
-------
TABLE 3.3. Actinides in the Fuel Charged to the PWR with
Self-Genera ted Pu Recycle (1000 !%e, with U and
Pu Recycle^/)
1. 3% 235U make-up fuel
Radionuclide
Urani urn-'
235
238
Total
kg/yr
5.97 x 10
1.93 x 10
+02
+04
1.99 x 10
+04
Ci/yr
1.28
6.43
7.71
Weight %
3.00
97.00
100.00
c/
2. Recycled plutonium fuel-'
Radionuclide
Urani um-
235
238
Total
5.40 x 10
7.49 x 10
+01
+03
7.54 x 10
,+03
1.16 x 10
2.50
2.62
-01
0.71
99.29
100.00
Plutonium^ 239
240
241
242
2.66 x 10
1.98 x 10
1.17 x 10
8.40 x 10
+02
+02
+02
+01
6.65 x 10
+02
1.63 x 10
4.49 x 10
1.32 x 10
3.28 x 10
+04
+04
+07
+02
a = 6.15 x 10
6 = 1.32 x 10
+04
+07
40.00
29.78
17.59
12.63
100.00
-'30.4 Mw-day/kg HM, 34.2% thermal efficiency, 80% capacity factor,
near-equilibrium cycle.
-/23ltU is not included.
-'150 days cooling of discharged fuel before reprocessing. 1.5%
loss in reprocessing, 1.5% loss in fabrication. Natural uranium
is added to the recycled plutonium to dilute the recycle fuel to
proper enrichment.
d/236pu anc| 238pu are not included.
3-6
-------
TABLE 3.4. Actinides in the Fuel Discharged From The
PWR with Self-Gene rated Pu Recycle!/
(1000 Mwe, with U and Pu Recycle^/)
1. 3% 235U make-up fuel
Radionuclide kg/yr Ci/yr Weight %
Uranium^' 235
236
238
Total
Plutonium^ 239
240
241
242
Total
1.59 x 10+2
7.50 x 10+1
•4-A.
1 .89 x 10+4
1.91 xlO+04
9.90 x 10+01
4.10 x 10+01
2.50 x 10+01
9.00
1.74 x 10+02
3.41 x 10"1
4.76
6.28
a =1.14 x 10+01
6.07 x 10+03
9.29 x 10*03
2.81 x 10+06
3.51 x 10+01
a = 1.54 x 10+04
0.83
0.39
98.78
100.00
57.39
23.05
14.43
5.13
100.00
3 = 2.81 x 10+06
2. Recycled plutonium fuel
Radionuclide
Uranium2/ 235
Plutoni
236
238
Total
urn^/ 239
240
241
242
Total
3.40 x 10+01
4.00
7.34 x 10+03
7.38 x 10+03
1.85xlO+02
1.70xlO+02
9.80 x 10+01
8.20 x 10+01
5.35 x 10+02
7.29 x 10"02
2.54 x 10"01
2.45
a = 2.78
1.13xlO+04
3.85 x 10+04
1.10xlO+07
3.20 x 10+02
a - 5.01 x 10+04
0.47
0.06
99.47
100.00
34.42
31 .83
18.31
15.44
100.00
3 = 1.10 x 10+07
-/ immediately after discharge
-/ 30.4 Mw-day/kg HM, 34.2% efficiency, 80% capacity factor,
near equilibrium fuel cycle
£/ 23fu 237u an(j 239y are not included
d/ 236pUj 238pu, and 21+3Pu are not included
O — /
-------
on fuel reprocessing so as to provide better foundation for future
facilities necessary to reprocess discharge fuel from breeders.
Similar reprocessing will also be necessary to obtain any significant
benefits from thorium fueling.
The uranium-plutonium recycle flowsheet of Figure 3.3 is calculated
for self-gene rated plutonium recycle. Alternatively, the plutonium re-
covered from fuel discharged from uranium-fueled reactors can be blended
with natural uranium to refuel another light-water reactor, as shown
in Figure 3.4. This uranium-plutonium fueled reactor is similar to
the uranium-fueled reactor in Figure 3.1, but it requires a larger
number of control absorbers because of the large neutron-reaction
cross section of fissile plutonium. Additional plutonium is formed in
this reactor during irradiation, and the plutonium in the discharge
fuel is recovered and recycled. The make-up plutonium for this cycle
very nearly equals that recovered from the fuel discharged by about
three uranium-fueled reactors operating as in Figure 3.2.
Such uranium-plutonium-fueled water reactors have been suggested
as piutoniurn-burner reactors to be located at centers where are also
co-located facilities for fuel reprocessing and for fabricating recycle
uranium-plutonium fuel. Discharge fuel from externally located
uranium-fueled reactors fueled with uranium or with uranium and
thorium would be sent to these centers for reprocessing, and the
recovered plutonium would be consumed on site in the plutonium-
burner reactors. Such centers have been proposed as a means of local-
izing the use of plutonium, thereby avoiding the safeguards issues
associated with shipping plutonium (Ul, HI). Examples of such inter-
national centers servicing off-site national reactors are given in
Chapter 7.
Although the uniformity of fuel charged to pi utoni urn-burner
reactors should lead to more optimum core loadings for these reactors
and greatest economy of plutonium utilization, the first plutonium
recovered for recycling is more likely to be returned as a partial
reload for the same reactor from which it was discharged. This is
the self-generated recycle operation of Figure 3.3. Approximately
one fourth of the reactor is fueled with natural uranium blended with
recycled plutonium, and three fourths is fueled with uranium enriched
to 3% 235U.
3.2 Thorium-Fueled Pressurized-Mater Reactors
The same pressurized-water reactor can be adapted to thorium
fueling, whereby natural 232Th replaces the function of the 238U
isotope in the previous flowsheets. The make-up fissile material
is either 93.5% 235U, as shown in Figure 3.5, or plutonium recovered
from the discharge fuel from uranium-fueled water reactors, as shown
3-8
-------
UD
1000 Mw
Pigford-Yang, 1977
Natural
Uranium
).7I5%235U
26.5 Mg
Plutoni
Fuel
Fabrication
i
um
— ta.
> t
i
«rt-
l"te
28.I
Mg
Light Water
Reactor
E=30.4 Mw day/kg
Fuel Life= 3yr
n =0.342
L=0,80
Plutonium Recycle
Fuel
Reprocessing
1
•
Fission
Products _
0.9!2Mg
Depleted
Uranium
71 % Fissile
0.70I Mg
54.5% Fissile
1.35 Mg
0.45%235U
25.5 Mg
Fig. 3.4 Lifetime-average annual quantities for PWR fueled with
Plutonium and natural uranium (E=fuel exposure, n=overall thermal
efficiency, L=capacity factor)
-------
1000 Mw
Pig ford-Yong, 1977
OJ
i
Thorium
24.6 Mg
93%*
0.501V
Natural
Uranium
1^.
lO? Mn
Fuel
Fabrication
55U
ig
Conversion
and
Isotope
Separation
25.6
Mg
Light Water
Reactor
E=334Mw day/kg
Fuel Life= 3yr
n = 0.342
L=0.80
Uranium Recycle
Mssion
_ Fuel Products^
Reprocessing 0.912 Mg
, , Thorium
r\—i x» t •
58% Fissile "'° my
0.896 Mg
L*-*~ Separative \Afork
108 Mg
10.25%259U
" 101 Mg
Fig. 3.5 Lifetime-average annual quantities for PWR fueled with
uranium and thorium (E=fuel exposure, n=overall thermal efficiency,
Incapacity factor)
-------
in Figure 3.6. The 233U, resulting from neutron absorption in
thorium, and other uranium isotopes are recycled. To simplify
comparison with the uranium-fueled PWR's discussed in Section
3.1, the same total heat generation per fuel rod for fueling with
urania or thoria has been assumed (SI). Because the thoria is
of lower density than urania, the average thermal exposure of
30.4 i% day/kg for urania fuel is equivalent on this basis to
33.4 Mw day/kg for thoria fuel. The recovered thorium is radio-
active because of 1.91-yr 228Th and "must be stored for several
years before it can be recycled (see Appendix A). Detailed data
on composition, of charge and discharge fuel for the near-equilibrium
fuel cycle are given in Tables 3.5 through 3.8. These data indicate
a concentration of 232U in the recycled uranium as high as 2600 ppm
for the near-equilibrium fuel cycle. This is 3.5 times greater
than the current estimate of 742 ppm for 232U in recycled bred
uranium for a near-equilibrium fuel cycle for the HTGR, as shown
in Chapter 5. Evidently the greater fuel lumping and close-
packed lattice of the PWR result in a higher flux of neutrons at
energies above the 232Th (n,2n) threshold.
3.3 Resource Requirements for Pressurized-Water Reactors
One purpose for considering thorium cycles in light water
reactors is to reduce the demands for uranium ore. The total
ore required to fuel a given reactor over its operating life must
include the ore to supply the start-up fuel inventory as well
as the annual replacement requirements accumulated over the
operating life. The lifetime ore requirements calculated for
the light-water reactor fuel cycles appear in Table 3.9. For
those cases involving recycle of fissile material, and/or supply of
fissile material recovered from fuel reprocessing, the reactor
is assumed to be fueled with slightly enriched (3%) uranium or
with 235U-Th fuel until sufficient fissile inventory is accumulated
within the reactor and fuel cycle so that the reactor can then
operate on the equilibrium fuel cycle. Therefore, for first-
generation recycle reactors there can be a considerable delay before
the resource advantage of recycle is manifested. For subsequent
reactors operating on the same fuel cycle, i.e. "second-generation"
reactors, the accumulated fissile inventory in the core and in
the fuel cycle from decommissioned reactors can be taken over so
that these new reactors can operate on the equilibrium fuel cycle
over their entire operating life, thereby achieving greater ore
savings. This assumes, of course, that at the time the "second-
generation" light-water reactors are to be constructed there are
no new types of reactors, such as breeders, which are more resource
efficient and which could better utilize the accumulated fissile
inventory.
It is apparent that thorium fueling in water reactors results
in only a small reduction in the uranium ore requirements, as
3-11
-------
1000 Mw
Pig ford-Yang, 1977
Thorium
24 I Mg
u>
ro
Fuel
Fabrication
Plutonium^
71 % Fissile
0.741 Mg
25.6
Mg
Light Water
Reactor
E=33.4 Mw day/kg
Fuel Life = 3yr
n=0.342
L=0.80
U - Pu Recycle
Fuel
Reprocessing
Fission
Products
0.9l2Mg
0417 Mgll, 58% Fissile
0.716 MgPu, 43.3% Fissile
Thorium
23.2Mg
Fig. 3.6 Lifetime-average annual quantities for PWR fueled with thorium,
Plutonium, and recycled uranium (E=fuel exposure, rpoverall thermal
efficiency, Incapacity factor)
-------
TABLE 3.5.
Actinides in the Fuel Charged to the U-Th
Fueled PWR (1000 Mwe, with U Recycle!7)
1. 93% 235U and thorium make-up fuel
Radionuclide
Thori urn 232
Total
Uranium 234
235
236
8.73 x 10
+03
8.73 x 10
1.90 x 10
3.67 x 10
9.00
+03
+01
+02
3.95 x 10
+02
9.55 x 10
-01
a = 9.55 x 10
1.18 x 10
7.87 x 10
3.00 x 10
-01
+02
-01
-03
a = 1.19 x 10
+02
Height %
100.00
100.00
4.70
93.00
2.30
100.00
2. Recycled uranium fuel-
Radionuclide
Thorium
Lira ni urn
232
Total
232
233
234
235
236
238
Total
kq/yr
1.53 x 10+04
1.53xlO+04
1.88
4.34 x 10+02
2.16 x 10402
1.7SxlO+02
3.14 x 10+02
3.30 x 10+01
1.17 x 10+03
Ci/yr
1.67
ct = 1.67
4.03 x 10
4.11 x 10
1.34 x 10
3.75 x 10
1.99 x 10
1.10 x 10
+04
+03
+03
-01
+01
-02
a = 4.58 x 10
+04
Meight %
100.00
100.00
0.16
36.97
18.47
14.91
26.75
2.81
100.00
-' 33.4 Mw-day/kg, 34.2% thermal efficiency,
near-equilibrium fuel cycle.
capacity factor,
—' 150 days cooling of discharged fuel before reprocessing, 1.5%
loss in fabrication. Natural uranium is added to the recycled plu-
tom'um to dilute the recycle fuel to the proper enrichment.
3-13
-------
TABLE 3.6. Actinides in the Fuel Discharged From the
U-Th Fueled PWRf!/
(1000 Mwe, with U Recycle-')
1. U and thorium make-up fuel
Radionuclide kg/yr
Thorium^
232
Total
8.50 x 10
+03
Protactinium^'
Total
Uranium-
/
232
233
234
235
236
238
Total
8.50 x 10
233 1.20 x 10
+03
+01
1.20 x 10
3.40 x 10
1.08 x 10
2.60 x 10
1.04 x 10
4.60 x 10
8.00
+01
,-01
+02
+01
,+02
,+01
2.92 x 10
+02
Ci/yr
9.30 x 10
-01
a= 9.30 x 10
2.49 x 10
-01
+08
B= 2.49 x 10
7.28 x 10
1.02 x 10
1.61 x 10
2.23 x 10
2.92
2.67 x 10
+08
+03
+03
+02
+01
-03
a =8.46 x 10
+03
Weight %
100.00
100.00
i
100.00
100.00
0.12
36.94
8.89
35.58
15.73
2.74
100.00
2. Recycled uranium fuel
Radionuclide kg/yr
Thorium^
232
Total
1.49 x 10
,+04
1.49 x 10
+04
Protactinium^
233 2.00 x 10
2.00 x 10
,+01
Uranium-'
232
233
234
235
236
238
Total
1.61
3.00 x 10*02
2.10 x 10+02
8.90 x 10*01
3.05 x 10*02
2.70 x 10401
9.33 x 10*02
Ci/yr
1.63
a = 1.63
4.15 x 10
+08
= 4.15 x 10
3.45 x 10
2.84 x 10
1.30 x 10
1.91 x 10
1.93 x 10
9.00 x 10
+08
+04
+03
+03
-01
+01
-03
a = 3.87 x 10
+04
Height %
100.00
100.00
100.00
100.00
0.17
32.17
22.52
9.54
32.70
2.90
100.00
- immediately after discharge
-/ 33.4 Mw-day/kg HM, 34.2% thermal efficiency, 80% capacity factor, near-
equilibrium fuel cycle. Np, Pu, Am, Cm are not included.
- Trace quantities of other thorium isotopes are not included.
— Trace quantities of other protactinium isotopes are not included.
-' 237U and 239U are not included.
3-14
-------
TABLE 3.7. Actinides in the Fuel Charged to the Pu-U-Th
Fueled PWR (1000 M*e, with U and Pu Recycle^)
1. Plutonium and thorium fuel
Radionuclide
Thorium 232
Total
Make-up h/
Plutonium2/ 239
240
241
242
Total
Recycled h/ ,
Plutonium^'' ^'239
240
241
242
kg/yr
1.13 x 10
+04
1.13 x 10
+04
,+02
3.02 x 10
1.21 x 10
7.60 x 10+01
2.70 x 10401
5.26 x 10
2.00 x 10
4.12 x 10
2.22 x 10
2.29 x 10
+02
+02
+02
+02
1.06 x 10
+03
2. Recycled uranium fuel-/' -/
Ci/yr
1.24
1.24
1.85 x 10
2.74 x 10
8.55 x 10
1.05 x 10
+04
+04
+06
+02
a = 4.60 x 10
B = 8.55 x 10
1.23 x 10
9.33 x 10
2.50 x 10
8.93 x 10
+06
+04
+04
+07
+02
Radionuclide
Thori urn
Urani urn
232
Total
232
233
234
235
236
Total
1
1
1
3
1
3
1
5
.21
.21
.49
.75
.44
.40
.70
.71
kg/yr
x
x
x
x
x
x
x
10+04
10
10
10
10
10
10
+04
+02
+02
+01
+01
+02
a = 1.06 x 10
6 = 2.50 x 10
Ci/yr
+05
,+07
1.32
a = 1.32
3.19 x 10
3.55 x 10
8.91 x 10
7.29 x 10
1.08
+04
+03
+02
+02
a = 3.63 x 10
+04
Weight %
100.00
100.00
57.42
23.00
14.45
5.13
100.00
18.82
36.76
20.88
21.54
100.00
Weight %
100.00
100.00
0.26
65.62
25.20
5.95
2.97
100.00
capacity factor,
-1 33.4 Mw-day/kg HM, 34.2% thermal efficiency,
near-equilibrium fuel cycle.
—' 236Pu and 238Pu are not included.
£ 150 days cooling of recycled fuel before reprocessing. 1.5% loss
in reprocessing, 1.5% loss in fabrication.
— thorium is added to the recycled uranium to dilute the recycle fuel
to proper enrichment.
3-15
-------
TABLE 3.8. Actinides in the Fuel Discharged from the
Pu-U-Th Fueled PWRf/
(1000 Mwe, with U and Pu recycle^/)
1. Plutonium-thorium fuel
Height*
100.00
100.00
100.00
100.00
0.20
92.23
6.19
1.38
100.00
18.80
38.69
20.99
21.52
100.00
Weight
100.00
Radionuclide
ThoriumC/ 232
Total
Protactiniums!/
Total
Uranium?/ 232
233
234
235
Total
Plutoniumf/239
240
241
242
Total
2. Recycled
Radionucl ide
Thorium?/ 232
Total
Protactinium^.''
Total
Uranium^/ 232
233
234
235
236
Total
kg/yr
1.11xlO+04
1.11xlO+04
9.00
9.00
2.90x10-01
1.34xlO+02
9.00
2.00
1.45x10+02
2.06x10+02
4.24x10+02
2.30x10+02
2.36x10+02
1.10x10+03
uranium fuel
kg/yr
1.1 7x1 0+04
1.1 7x1 0+04
233
1.70x10+0'
1.70xlO+01
1.25
2.35x10+02
1.47x10+02
3.60x10+01
2.30xlO+01
4.42xlO+02
Ci/yr
1.22
oc= 1.22
5.12x10+05
B = 5.12x10+05
6.21x10+03
1.27x10+03
5.57x10+01
4.29x10-03
a= 7.36x10+03
1.26x10+04
9.60xlO+04
2.59xlO+07
9.20x10+02
a = 1.10x10+05
B = 2.59xlO+07
Ci/yr
1.28
a = 1.28
3.53xlO+08
e = 3.53xlO+08
2.68xlO+04
2.23x10+03
9.10x10+02
7.72x10-02
1.46
a= 2.99xlO+04
100.00
100.00
100.00
0.28
53.14
33.24
8.14
5.20
100.00
a/ immediately after discharge
b/ 33.4 Mw-day/kg HM, 34.2% thermal efficiency,
near equilibrium fuel cycle
c/ Trace quantities of other thorium isotopes are not included.
d/ Trace quantities of other protactinium isotopes are not included.
e/ 236y, 237y and 238.j are not included
f/ 236pu and 238 Pu are not included
capacity factor,
3-16
-------
TABLE 3.9 30-Year Lifetime Ore Requirements for Pressuri zed-Water
Reactors (1000 Mwe Electrical Power, 80% Capacity Factor)
Natural Uranium^ Natural Uranium
Fuel
Cycle
(a) No recycle
(b) U recycle
Conversion
Ratio
0.60
0.60
0.2%
depleted U
4940
4070
0.25%
depleted U
5370
4487
relative
to , b/
no recycle-'
1.00
0.84
Thorium ,
Metric tons (Mg)-'
(c) U-Pu self-
generated
recycle
(d) 235U-Th,
U-re cycle
(e) Pu-Th, U-Pu
recycled^/
(f)
Second
generation
Pu-Th, U-Pu
recycle^/•
0.61
0.66
0.61
0.61
3340
2810
3250
2560
3680
3060
3584
2790
0.68
0.57
0.67
0.52
774
157
220
-' Uranium contained in U,0q concentrate. To obtain short tons of
U,0Q multiply by 1.297.
O O
- Calculated for 0.25% depleted U from isotope separation.
— Thorium contained in ThO? concentrate. To obtain short tons of
Th02 multiply by 1.2517.
— Includes U-fueled water reactor to supply make-up Pu. Total system
operates for 30 yr at 1000 Mw.
e/
- Starts with equilibrium reactor and fuel-cycle fissile inventory.
3-17
-------
compared with the analogous case of uranium fueling with recycle
of uranium and plutonium. The greatest ore saving per unit of
total generated electrical energy for first-gene ration thorium
fueling results from the 235U-Th system of Figure 3.5. However,
the use of fully enriched (93%) 235U is subject to special safeguards
concerns, as is discussed in Chapter 6. If a stockpile of discharge
fuel from uranium-fueled water reactors were accumulated and then
later reprocessed, use of the recovered plutonium to fuel the thorium
flowsheet of Figure 3.10 could be a more direct way of initiating
thorium fueling. Variations of these thorium cycles which have
been suggested to mitigate the safeguards issues are discussed in
Chapter 6.
Recent estimates (SI) indicate about the same fuel-cycle costs
with Pu-Th fueling as for U fueling with U-Pu recycle. 235U fueling
is estimated to be more expensive, because of the relatively high
costs of 93% 235U. However, there is too much uncertainty in the
cost of fuel reprocessing, particularly for thorium fuel reprocessing,
for the small differences in the estimated costs for these fuel
cycles to be significant. Also, Thorex reprocessing technology has
not been brought to the commercial scale of Purex reprocessing, and
additional costs of development can be expected (HI, D5).
Assuming no significant cost advantage for thorium fueling in
the near future, thorium fueling could become more attractive at a
future time when uranium supplies are more limited and the cost of
uranium is relatively high. Then reactors with less uranium con-
sumption would have a greater economic advantage and would be more
useful to the power economy. There would also be greater incentive
to redesign light water reactors to higher conversion ratios for
better ore utilization than is indicated in Table 3.9.
The proposed light-water breeder reactor is such a concept. It
involves a higher ratio of fuel to water than in present reactors,
separated and localized regions of fissile and fertile material, and
the use of moveable fuel for reactivity control. All of these modi-
fications increase neutron absorption in thorium, resulting in
higher conversion ratio. Thorium-cycle conversion ratios of near
unity seem achievable. However, the breeding gain is very small and
specific power is low, so pre-breeders of intermediate conversion
ratios are proposed as a means of providing the start-up fissile in-
ventory. Analyses have indicated an increased ore requirement
during the period of introducing prebreeders and breeders, and net
ore savings only after a very long period of operation. Thus, the
value of the light-water breeder is marginal relative to other alter-
natives.
3-18
-------
A possibly more useful and realistic concept, to improve the
conversion ratio and ore consumption with near-term light-water
reactors, is to modify these reactors for "spectral shift" operation
(B2). The reactor coolant system could be modified so that heavy
water (D^O) could be introduced into the coolant at controlled con-
centrations. After each refueling cycle the excess neutron produc-
tion from fresh fuel would be controlled by replacing enough H?0
with DpO for less efficient neutron moderation. This excess neutron
production, normally absorbed in boron or other non-fertile absorbers, or
would be consumed by the absorption resonances of the fertile materials
238U or 232Th,. thereby increasing the fissile production and conver-
sion ratio.
As fuel burnup proceeds the D20 is replaced by H20 to maintain
reactivity, and the process is repeated for each refueling cycle.
Typically, at the beginning of a refueling cycle the reactor coolant
would consist of about 75% DpO. During the cycle the coolant is di-
luted with normal water, resulting in a concentration of less than
2 to 5% D20 at the end of the one-year cycle. A facility must
be provided to reconcentrate the heavy water. The spectral shift
reactor received some attention over a decade ago, but it was not
justified economically at that time. Preliminary estimates (SI)
indicate significant improvement in conversion ratio over that of
any of the light water cycles listed in Table 3.3, even using the
lattice of present-day pressurized water reactors. For fuel burnups
of 33.4 megawatt day per kg a 235U-Th loading was calculated to
operate at an integral conversion ratio of 0.7 and a 233U-Th loading
at a conversion ratio as high as 0.87. This indicates the possibility
of reducing the lifetime ore requirements well below 2300 metric tons
of natural uranium.
Benefits from the higher conversion ratios of spectral shift
LWR's must be balanced against increased costs resulting from the
increased complexities of using heavy water in LWR's. Facilities
must be provided to adjust DpO concentration in the LWR coolant and
to re-enrich the D20 diluted by H20 during the fuel cycle. An on-
site distillation system for heavy-water enrichment is a possibility.
The presence of concentrated deuterium in the coolant will
increase the rate of production of tritium in the coolant. In
the pressurized water reactor this increase in tritium production by
neutron absorption in deuterium will be offset, in part, by the loss
of tritium production from fast-neutron reactions in dissolved boron,
since dissolved boron will no longer be needed for reactivity control.
A higher concentration of tritium in the coolant may complicate the
open-core refueling techniques now used in LWR's, because of the
possibility of tritium escape during refueling. It may also require
additional controls to minimize the environmental release of tritium
via non-condensable off gases during normal operation.
Control of burnup reactivity by spectral shift boiling-water
reactors would eliminate the burnable-poison absorbers now incorpor-
3-19
-------
ated in the fuel rods in these reactors. However, since the burnable
absorbers also provide an effective means of adjusting the axial dis-
tribution of neutron flux and power density in these reactors, some
other technique must be developed for power-density control in a
spectral-shift boiling-water reactor. The larger negative void co-
efficient of reactivity in spectral shift operation would also be a
problem.
Nevertheless, the spectral shift concept might be relatively easy
to implement in some present PWR's and should be included in further
evaluations of alternatives for improved resource utilization.
Spectral shift operation with thorium fueling provides the greatest
gain in resource utilization, but the improved utilization of uran-
ium fuel with spectral shift operation may become justified when
uranium ore prices increase.
There is another way in which thorium may be utilized in a present-
day or spectral shift light-water reactor, not merely to extend resources
but as a part of an overall approach to international safeguards. As
is discussed more completely in Chapter 7, the recycled uranium in a
uranium-thorium cycle is diluted with 238U to about 15 to 20% fissile
isotopic concentration. This results in somewhat less plutonium pro-
duction than in a low-enrichment uranium cycle. Such a fuel cycle
is completed by storing or reprocessing the discharge fuel to recover
the uranium and plutonium.
In summary, the use of thorium in present light water reactors
offers a real but marginal advantage for resource extension alone.
Larger benefits are possible with redesign of the reactor core or of
the moderator-coolant system. Were the fast breeder reactor to be delayed
or eliminated altogether, it might be desirable to introduce the use
of thorium in LWR's modified for higher conversion ratios, since the
overall reduction in uranium ore demand for a larger number of LWR's
could be important.
3-20
-------
4- Fuel Cycles for Heavy-Water Reactors
4.1 Uranium Fueling as a Reference Case
The Canadian (CANDU) version of the heavy water reactor is being
considered as a possible means of better resource utilization in the
U.S., if such conservation should become necessary because of delays
in the breeder program. The flowsheet for the natural-uranium version
of the CANDU reactor (Fl, Ml) is shown in Figure 4.1. Greater quanti-
ties of plutonium are present in the total discharge fuel from this
reactor than in the case of a light water reactor of the same power,
because of the higher conversion ratio and shorter fuel irradiation
exposure of the heavy water reactor. However, the large throughput of
uranium results in a relatively low concentration of plutonium in the
discharge fuel. Present costs of uranium and of fuel-cycle operations
do not now justify reprocessing to recover the plutonium from the
fuel discharged from these heavy water reactors, so the discharge
fuel is now put into long-term storage. However, future higher
costs of uranium ore may ultimately justify reprocessing the fuel
to recover and recycle plutonium. As shown in Table 4-1, the ore
utilization of the non-recycle CANDU is 25% better than that for the
uranium-fueled PWR without recycle.
Even without fuel reprocessing, the ore consumption of the CANDU
reactor can be reduced by fueling with slightly enriched uranium, as
illustrated in Figure 4.2. By increasing the enrichment to 1.2%,
the average fuel exposure is increased from 7.5 Mw day/kg to 21 Mw day/kg
(B3, Tl). Because of the greater burnup, the consumption of natural
uranium is only 71% of that of the natural-uranium-fueled CANDU
reactor. The concentration of plutonium in the discharge fuel
increases, but the total amount of plutonium in the discharge fuel
is only 30% of that from the natural-uranium-fueled CANDU reactor.
The calculated operation of the heavy-water reactor with natural
uranium and self-gene rated plutonium recycle (B 3) is illustrated in
Figure 4.3. Recycling the plutonium makes a significant difference
in the fuel burnup, which rises from the low value of 7.5 Mw days/kg
for natural uranium to 18 Mw days/kg. The data in this flowsheet were
derived from calculations which assumed that the plutonia-urania fuel
with the same fuel and cladding dimensions as the present CANDU fuel
can operate to the higher burnups without modification. This is an
optimistic assumption, since the higher barnups will generate more
fission gases. Fission-gas plenums and thicker fuel cladding may be
required. As shown in Table 4.1 the lifetime uranium ore requirements
for the CANDU reactor with self-generated plutonium recycle are about
two-fold less than for the present non-recycle operation with natural
uranium. Although for near-term ore costs the burnup per cycle is
still too low for reprocessing and recycle to be more economical than
the non-reprocess ing stowaway cycle, at some future higher price of
4-1
-------
1000 Mw
TPigford. 8/76
Natural
Uranium
0.7I5%H5U
131 Mg
Fuel
Fabrication
Heavy Water
Reactor
E = 7.5 Mw Day/Kg
Fuel Life = I yr
n = 0.305
L=0.80
Fuel
Storage
O.I7%235U
0.27% Fissile Pu
ro
Fig. 4.1 Annual quantities for natural-U-fueled CANDU
reactor (E=fuel exposure, n=overall thermal efficiency,
Incapacity factor)
-------
1000 Mw
CO
Fuel
Fabrication
1.2 % 235U
46.3 Mg
Natural
Uranium^
35
U
95.5 Mg
0.71%
Heavy Water
Reactor
E = 2I Mw Day/Kg
Fuel Life =2.8yr
fj* 0.305
L=0.80
Conversion
and Isotope
Separation
0.25%235U
48.3 Mg
Fuel
Storage
0.08
0.30% fissile Pu
Separative Work
27.1 Mg
Pig ford • Yang
Fig. 4.2 Annual quantities for slightly enriched U-
fueled CANDU reactor (E=fuel exposure, n=overall thermal
efficiency, Incapacity factor)
-------
1000 Mw
Pigford - Yonq
Natural
Uranium
07I5%235U
53.2 Mg
Fuel
Fabrication
Heavy Water
Reactor
E =18 Mw Day/Kg
Fuel Life = 2.4yr
rj =0.305
L = 0.8
Plutonium Recycle
0.214 Mg fissile Pu
Fuel
Reprocessing
Fission
Products
1.013 Mg
Depleted U
51.7 Mg
Fig. 4.3 Annual quantities for equilibrium U-fueled
CANDU reactor, with self generated Pu recycle (E=fuel
exposure, n=overall thermal efficiency, Incapacity factor)
-------
TABLE 4.1. 30-Year Lifetime Ore Requirements for Heavy Water Reactors
(1000 Md electrical power, 80% capacity factor)
Fuel Cycle
(a) Natural U fuel, no recycle
(b) 1.2% 235U in U, no recycle,
0.20% depleted U
0.25% depleted U
(c) Natural U fuel, Pu recycle
(d)
235U-Th fuel, U recycle
0.20% depleted U
0.25% depleted U
(e) Pu-th fuel, U-Pu recycle-'
d/
Conversion
Ratio
0.75
0.64
0.74
0.92
0.92
Natural .
Uranium -'
Metric tons
(Mg)
4060
2723
2860
2200
1303
1442
1765
Natural Uranium
Relative to
U/fueled PWR
b/
without reprocessing —
0.75
0.50
0.-53
0.41
0.24
0.27
0.33
Thorium -r
Metric tons
(Mg)
1126
616
- Uranium contained in U000 concentrate.
J O
To obtain short tons of
U,0Q multiply by 1.297.
h 7
- For 0.25% depleted uranium from isotope separation.
- Thorium contained in Th09 concentrate. To obtain short tons of
ThO, multiply by 1.2517. ^
t I L.
- Includes U-fueled CANOU to supply make-up. Total system power - 1000
4-5
-------
uranium ore such reprocessing fuel cycles could become economically
attractive.
4-2 Thorium-Fueled Heavy-Water Reactors
The same CANDU reactor can also be fueled with thorium and make-up
fissile material derived from an external source (B4, C2, Kl, Tl, T2).
Figure 4.4 is the flowsheet for the equilibrium cycle of the CANDU
reactor fueled with 93.5% 235U, thorium, and recycled uranium! The
fuel burnup has been specified at 27 megawatt days per kilogram of
heavy metal, near that typical of light-water fuel. As shown in
Table 4.1, uranium-thorium fueling increases the average conversion
ratio to 0.92, a result of the greater number of fission neutrons
per absorption for the bred and recycled 233U. The uranium ore con-
sumption is 39 to 45% less for this cycle than for uranium fueling
with self-generated plutonium recycle. Alternatively, the make-up
fissile material for the thorium-fueled CANDU reactor can be plutonium
recovered from uranium fuel discharged from a CANDU reactor or a
light water reactor. The flowsheet for this cycle at equilibrium,
utilizing pi utoni urn produced in a natural uranium CANDU reactor, is
shown in Figure 4.5. The lifetime ore requirements for the piutoniurn-
thorium CANDU reactor, shown in Table 4.1, are calculated for
30x0.8 Qw yr of electrical energy from a reactor system consisting
of a uranium-fueled CANDU reactor to provide the start-up and make-
up plutonium and a piutoniurn-thorium fueled CANDU to consume the
plutonium. The uranium ore required for this system is 20% less
than for the CANDU fueled with natural uranium and self-gene rated
plutonium recycle, and it is 22 to 36% greater than for the 235U-Th-
fueled CANDU with uranium recycle.
In the present conceptual design of a thorium-fueled CANDU
reactor the lattice spacing and specific power have been kept the
same as for the natural-uranium CANDU reactor. Because the fuel
burnup chosen for these thorium cycles is 3.6 times greater than
for present CANDU uranium fuel, the void volume in each fuel rod has
been increased by 9% to provide for the accumulation of fission gases (Tl).
The cladding dimensions have been kept the same, although the higher
burnups may require thicker cladding. There are no published data on
the performance of CANDU fuel elements at these high burnups.
The lifetime ore requirements for the CANDU with recycle are
significantly less than the ore requirements for any of the light water
reactor fuel cycles shown in Table 4.1. This is indicative of the
overall higher conversion ratio of the heavy water reactor. However,
the ore savings and savings in separative work must be balanced against
the higher fabrication and reprocessing costs resulting from the lower
fuel burnup of the heavy water reactor, differences in construction
costs, and the cost of heavy water.
4-6
-------
Thoriurp
34.6 Mg
Fuel
Fabrication
93.5%235U
0.226 Mg
Natural
Uranium
07I5%235U
45.3 Mg
Conversion
and
Isotope
Separation
1000 Mw
I
T. Pig ford, 1977
Heavy Water
Reactor
E = 27 Mw Day/Kg
Fuel Life = 3.6 yr
tf = 0.305
L=0.8
Fuel
Reprocessing
Uranium Recycle
0.434 Mg 233U, 0.078 Mg235U
0.966 Mg U
Separative Work
49.0 Mg
45.1 Mg
235,
Fig. 4.4 Annual quantities for equilibrium U-Th-
fueled CANDU reactor, with U recycle (E=fuel exposure,
n=overall thermal efficiency, Incapacity factor)
Fission
Products
1.013 Mg
Thorium
33.3 Mg
-------
-Pi
00
1000 Mw
T.Pigford,l977
\
Natural
Thorium
34. 8 Mg
Fissile
Plutoniun
Fuel
Fabrication
i
i
0.232 Mg
i
Heavy Water
Reactor
E = 27 Mw Day/Kg
Fuel Life = 3.6yr
n = 0.305
L=0.80
Uranium Recycle
Fuel
Reprocessing
\
0.427Mg233U
•
Fission
Products
1.013 Mg
Thorium
33.6Mg
0.042
0.673 MgU
Fig. 4.5 Annual quantities for equilibrium Pu-Th-
fueled CANDU reactor, with U recycle (E=fuel exposure,
n=overall thermal efficiency, Incapacity factor)
-------
A CANDU-type heavy water reactor can be modified to "operate at yet
higher conversion ratios, even as a thermal breeder with thorium make-up
and urani-um recycle (B3, B4, C2, Kl, Tl, T2). A conversion ratio of unity
can theoretically be obtained in the present CA.NOU lattice if fueled with
thorium at low fuel burnup. The conversion ratio can also be increased by
increasing the thorium loading, operating at lower specific power, and
increasing the calandria lattice spacing. Although the ultimate fuel
savings from the higher conversion ratios are ultimately very large, they
cannot be realized at the beginning of the reactor lifetime. A higher
initial fissile loading is required, and the smaller cumulative ore require-
ments are realized only after many years of operation.
The startup of the CANDU thorium breeder has been studied by Banerjee,
et al. (86, Bll). Prebreeding is initiated by fueling the CANDU reactor
with thorium and with plutonium recovered from irradiated CANDU uranium.
The discharged thorium-piutonium fuel is reprocessed, and recovered plutonium
and bred uranium are recycled with additional make-up or recycled thorium.
The cycle is subsequently repeated. Additional plutonium is added to each
recycle loading to maintain reactivity, but the necessary amount of plutonium
decreases with each subsequent cycle, until the reactor finally becomes
self-sustaining on thorium and recycled bred uranium.
For reactors with characteristics of the present CANDU reactors, self-
sustaining thorium breeding occurs at fuel exposures of 12.1 Mw day/kg.
The breeding ratio is unity, so all subsequent breeders must be started up
by the same process of utilizing piutonium recovered from irradiated uranium
fuel, or, alternatively, obtaining enriched uranium from isotope separation.
Based upon startup with CANDU plutonium, and using the cycle-by-cycle data
of Banerjee and Barclay (Bll) for Pu-Th fueling, with Pu-U recycle, the
total, plutonium required to achieve self-sustaining breeder is 4410 kg for
a 1000 Mw CANDU. The total amount of natural uranium to produce this amount
of plutonium in a CANDU is 1650 Mg.
The time-dependent demand for U uranium ore for startup of a 1000 Mw
CANDU thorium breeder is shown in Figure 4.6. One approach is to obtain an
initial amount of plutoniurn of 1962 kg, sufficient for a Pu-Th initial
loading. To produce this plutonium, 734 Mg of natural uranium would be
required to fuel a natural-uranium CANDU reactor. This corresponds to the
initial ore requirement of curve 1 of Figure 4.6. In subsequent cycles
additional plutonium is required, so additional uranium must be irradiated
in separate CANDU reactors to produce that plutonium. Assuming two years for
reprocessing and fabrication of recycled fuel,.the reactor becomes self-
sustaining in 10.8 years. This scenario is realistic when there exists a
stored inventory of irradiated uranium fuel to be reprocessed for the
startup plutonium inventory, or when there are existing uranium-fueled CANDU
reactors to produce the plutonium needed for thorium breeder startup.
If an inventory of irradiated uranium is not available, or if uranium-
fueled CANDU reactors are not available, the CANDU reactor destined to become
a thorium breeder can be operated initially on uranium fueling for 12.6 years,
at which time sufficient plutonium has been produced to convert the entire
reactor to Pu-Th fueling. Self-sustaining breeding occurs after 23.4 years.
The demand for uranium ore for startup of CANDU thorium breeders can be
spread over a longer period of time by starting the CANDU with uranium fueling
4-9
-------
2000
-51500
1 I
Pigford-Yang, 1973
I. Reactor is fueled with an initial inventory of Pu and Th.
Additional Pu is obtained from U-fueled CANDU reactors.
2. Reactor is fueled with U until sufficient Pu is made to
convert completely to Pu-Th.
3. Reactor is fueled initially with U and recycles Pu and
bred U with Th when available.
1
1
Figure 4.6
10 15 20 25 3O
Reactor Operating Time, years
Cumulative Requirement of Natural Uranium for the Pu-Topped,
Thorium-Fueled Self-Sufficient CANDU reactor (1000 Mwe, fuel
exposure at equilibrium = 12.1 Mwday/kg)
-------
and recycling the self-generated plutonium and bred uranium as Pu-Th-U fuel,
as it becomes available. In each subsequent cycle some of the reactor
pressure tubes previously fueled with natural uranium are converted to
thorium fueling. Thorium fuel elements progress through the calandria
tube more slowly than do the uranium fuel elements, because of the higher
burnup for thorium breeding. After 51 years the reactor becomes a self-
sustaining thorium breeder, with 99% of the pressure tubes operating on
thorium fuel.
The total ore requirement for breeder startup is the same for each of
the three different approaches considered above.
As seen above, to produce the total plutonium required to start the
thorium CANDU breeder requires the operation of 1000 Mwe uranium-fueled
CANDU reactor for 12.2 years on natural-uranium fuel, consuming 1650 Hg
of natural uranium. If, instead, the plutonium recovered from the discharge
uranium fuel were recycled as mixed-oxide fuel with natural uranium, the ore
consumption of this original CANDU reactor for 12.2 years could have been
reduced to 895 Mg of natural uranium, a saving of 800 Mg. Therefore, the
uranium ore directly attributable to breeder start-up is 800 Mg. This
means that 800 Mg of additional natural uranium would be required in the
original CANDU reactor if the plutonium in the discharge fuel is to be
accumulated for later breeder startup, instead of utilizing the plutonium,
when available, in self-generated recycle to reduce the consumption of
natural uranium.
4-11
-------
5- High-Temperature Gas-Cooled Reactor
5.1 Reactor Characteristics
The high-temperature gas cooled reactor (HTGR) moderated with
graphite and cooled with helium, is undergoing demonstration tests
as an alternative nuclear power plant of the future. As shown in the
overall material flowsheet of Figure 5.1 for the HTGR reference
design, the reactor is fueled with thoeium, make-up 235U, and
recycled uranium. The overall flowsheet of Figure 5.1 is a
simplified composite of the more detailed flowsheet of Figure 5.2,
which shows the several different and segregated fuel streams,
scrap recycle (Jl), and process losses.
The fuel consists of coated particles of uranium and thorium
embedded in a prismatic graphite matrix. Helium coolant flows
through holes in the graphite. The fuel-moderator prisms, of
hexagonal cross section, are stacked to form the core structure.
The graphite matrix provides a means of obtaining very high fuel
burnup without loss of mechanical integrity. The current design
burnup is 94.3 megawatt days per kilogram of uranium and thorium,
which is about three times that experienced in light water reactors.
A commercial prototype is now operating at Fort St. Vrain in
Colorado. However, the U.S. manufacturer, General Atomic, has
recently withdrawn its earlier sales of full-scale commercial
plants. Development of HTGR fuel reprocessing and refabrication
technology continues under DOE sponsorship. It is uncertain
when and if this reactor will return to the commerical U.S. market.
Although the overall HTGR fuel cycle appears similar to the
235U-thorium fuel cycles for the light-water reactor (Figure 3.5)
and for the heavy-water reactor (Figure 4.4), the HTGR flowsheet
differs in detail because of the plans to segregate the various
fissile and fertile fuel streams for the purpose of reprocessing and
recycle. This is shown in the more detailed flowsheet of Figure
5.2. The reprocessing flowsheet further differs from that for the
thoria fuels of the light-water and heavy-water reactors because
of the large quantity of graphite in the HTGR fuel matrix. In
HTGR fuel reprocessing this graphite is burned to expose the fuel
particles for acid dissolution.
The discharge fuel is processed to recover the uranium remaining
from the initial make-up 235U, which is then recycled for one more
pass through the reactor. Also recovered for recycling is the
uranium, largely fissile 233U, formed by neutron-capture reactions
in thorium.
Corresponding material quantities for the
near-equilibrium fuel cycle, derived from data for the ninth
reload (H4), are given in Tables 5.1, 5.2, 5.3, and 5.4. The
actinide quantities in the composite of the three discharge-fuel
streams are given in Table 5.5.
5-1
-------
lOOOMw
tn
i
Thorium
7. 44 Mg
Fuel
Fabrication
j
93.5% 235U
0.386 Mg
Natural
Uranium
0.71 % 235U
-r~» O h4_
i
Conversion
and Isotope
Separatbn
«4-^
^
Gas Cooled
Reactor
E = 95Mw Day/Kg
Fuel Life = 4 yr
>7 = 0. 387
L =0.80
Uranium Recycle
i
Reprocessing
0.167 Mg 233U
0.033 Mg 235 U UrQnium
Discard
^^ Separative Work 2.29 %235
82-6M9 0.058M?
Fission
Products
3 0. 798 Mg
Thorium
6.75Mg
(U
1
0.25%U
76.8 Mg
Fig. 5.1 Annual quantities for the near equilibrium U-Th-fueled
HTGR, with U recycle (E=fuel exposure, n=overall thermal efficiency,
Incapacity factor)
-------
Natural Uranium
%_ Kg
234U 0.006 4.63
235U 0.715 552
238U 99.279 76655
Total 100 77212
235
0.74
93.12 360
536 22.7
38U
Total 100 386
1000 Mw
Isotope
Separation
1
0.75 %L
2.90 K
Conversion
and
Fabrication
OSS I
gy.
en
co
% Kg
0.0018 1.38
0.25 192
99.75 76634
Total 100 76826
3953 Kg Th
74 43 Kg
Thorium »
(with 100 ppm 230Th)
29.6 KgThl
230Th'
Kg
232
234
Th
U
0.072
726
121 1.07
235U 29.52 26.2
2BJ 5227 46.3
238U 17.00 15.1
Total 815.2
Th02
microsphere
preparation
Pigford -Yang
0.75% Loss
55.8 KgTh
3434 Kg Th ^
Gas Cooled
Reactor
E*95MwDoyKg
Fuel Life - 4 yr
•- 0 387
= 080
I Year;
Storage
232
232
233l
234l
2351
236,
Th
U
--K9—
026
2615
0.04 0.09
79.48 166.65
16.52 36.64
338 7.09
0.57 1.19
2825
Total Ztto 1512 Kg U
29% U Scrap Recovery. 58 Kg Th
233U Recycle
_L
Conversion..
i and
Fabrication
I % Loss
5.22 Kg U
34.3KgTh,
235U Recycle
0.75%
U and Th Losses
3.94 Kg U
510 KgTh
Actinides and
Fission Products
Kg C[
Pa 0.0015
Np 10.90
Pu 8.43
Am 0.19
Cm 0.07
FR 798
3.03xl04
7.70
I.57»K)5
147.9
l.22«04
3.87»I07
Depleted Uranium
Kc
Total
Fig. 5.2 Detailed annual mass flow sheet for the near-equilibrium
235U-Th-fueled HTGR, with U recycle (E=fuel exposure, n=overall
thermal efficiency, L=capacity factor)
-------
1.
TABLH 5.1
Actinides in the Fuel Charged to the U-Th Fueled HTGR
(1000 Mwe, high-temperature gas-cooled reactor with uranium recycle- )
235U-Th make-up fuel-''
Radionuclide
Thorium 230
232
Total
kg/yr
Ci/yr
3.89x10
,-01
3.92x10
,+03
Uranium
234
235
236
238
Total
3.92x10
2.84
3.58x10"
1.09
2.25x10*
3.84x10*
.+03
•02
2. Once recycled 235U-Th fuel-'', -f
Radionuclide
Thorium 230
'232
Total
Uranium 234
235
236
238
Total
kg/yr
7.20x10
7.26x10'
,-02
+02
7.26x10
1.07
2.62x10'
4.63x10'
1.SlxlO
+02
,+01
,+01
,+01
8.87x10
3. Recycled 233U-Th fuel^/' £
+01
Radionuclide
Thorium 230
232
Total
Uranium 232
233
234
235
236
Total
kg/yr
2.59x10"
2.62x10'
+03
2.62x10
9.00x10
1.67x10'
3.46x10
7.09
1.19
2.10x10'''
+03
-02
,+02
+01
7.56
4.29x10
-01
7.99
6.66x10
7.67x10
6.82x10
7.50x10
+03
,-01
,+01
-03
6.73x10
Ci/yr
+03
1.40
7.94x10
-02
l.t
6.66x10'
5.61x10
2.94
5.02x10'
,+03
-02
-03
6.66X10*55
Ci/yr
5.04
2.86x10
-01
5.33
1.93x10'
1.58x10'
2.14x10
2.31
+03
+03
+02
6.82x10'
,+01
3.79x10'
+DT
100.00
0.74
93.12
0.28
5.86
100.00
0.01
99.99
100.00
1.21
29.52
52.27
17.00
100.00
0.01
99.99
100.00
0.04
79.49
16.52
3.38
0.57
100.00
- 95 Mv-day/kg HM, 38.71 thermal efficiency, 80% capacity factor, near-equilibrium fuel cycle.
—' Natural thorium is assumed to contain 100 ppm Th.
- 1 year cooling of discharged fuel before reprocessing, 0.75% loss in reprocessing, II loss in
fabrication. Thorium is added to the recycled uranium to dilute the recycle-fuel to proper enrichment.
5-4
-------
TABLE 5.2
Actinides in Discharged 235U-Th First Cycle Make-Up OTGR Fuel
(1000 Mwe U-Th fueled high-temperature gas-cooled reactor- )
Radionuclide
Fissile Fuel
Fertile Fuel
Thorium^ 228
229
230
231
232
234
Total
Protactinium 233
234
Total
Uranium^' 232
233
234
235
236
237
238
Total
Neptunium 237
Total
Plutonium 236
238
239
240
241
242
Total
Americium 241
242m
243
Total
Curium 242
243
244
245
Total
TOTAL ACTINIDES
kg/yr
1.26xlO"7
1.39xlO"8
1. 77x10" 5
1. 22x10" 10
4.55xlO"6
1.22xlO"12
2.24xlO"5
1.04xlO"12
1.42xlO"17
1.04xlO"12
9.43xlO"6
1.49X10"6
1.23
2.98X101
S.28X101
9.S4xlO"9
1.72X101
l.OlxlO2
6.46
6.46
2.52xlO"6
3.14
6. 82x10" l
3.17X10"1
_1
2.96x10 i
S.llxlO"1
4.75
2.03xlO"2
1.72X10"4
9.00xlO'2
l.lOxlO"1
1.13X10"3
7.45xlO"S
4.00xlO"2
1.68x10'*
4.14X10"2
1.12xl02
Ci/yr
l.fMxlO"1
2.?6xlO"6
3.44xlO"4
6.49xlO"2
4.98xlO"10
2.82xlO"5
a- 1.04X10"1
B= 6.49xlO"2
2. 16x10" 5
2.82xlO"8
6= 2.16xlO"5
2.02X10"1
r. 41x10" 5
7.61
6.39xlO"2
3.35
7. 79x10" *
S.73xlO"3
a= 1.12X101
B= 7.79X10"1
4.56
a- 4.56
1.34
5.49xl04
4.13X101
7.18X101
A
3.33x10
1.21
OF S.SOxlO4
6= 3.30xl04
6.58X101
1.67
1.67X101
a- 8.25X101
B- 1.67
3.74xl03
3.43
3.33xl03
2.64xlO"2
a- 7.07X103"
a- 6.22xl04
B= 3.33xl04
kg/yr
7.72xlO~4
8.37xlO~4
1.43X10"1
1.62X10"11
3.64X103
2.7SxlO"8
3.64xl03
-d
7.88x10
3.21xlO"13
7.88xlO"4
S.56xlO'2
9.72X101
1.94X101
3.95
6.26-10"1
4.71xlO"12
2.72xlO"5
1.21xlOZ
4.04xlO~2
4.04xlO'2
1.02xlO"12
9.83xlO"3
7.67xlO*4
2.40xlO'4
1.46xlO"4
5.40xlO'S
l.lOxlO*2
8.75xlO~6
2.78xlO"7
2.28xlO'5
3.18xlO'5
1.82xlO"7
5.56xlO"9
4.50xlO"S
2.52xlO"7
4.54xlO"5
Ci/yr
6.34xlOZ
1.78X10"1
2.77
8.62xlO"3
3.98X10"1
6. 36x10" l
a.' 6.37xlOZ
6= 6.45X10"1
A
1.64xl04
6. 36x10" 4
6= 1.64xl04
1.19xl03
9.21xl02
1.20xl02
8.47xlO"3
3.97xlO"2
3.85xlO"4
9.03xlO"9
a= 2.23xl03
B= 3.85xlO"4
2.85xlO"2
o= 2. 85x10" 2
5.43xlO"7
1.72xl02
4.70xlO"2
5.44xlO"2
1.65X101
2.11xlO"4
a= 1.72xl02
B- i.esxio1
2.84xlO"2
2.70xlO"3
4.22xlO"3
a= 3.26xlO"2
B= 2.70xlO"3
6.02X10"1
2.56xlO"4
3.75
3.9SxlO"S
a- 4.35
3.76x10°
a« 3.04xl03
B- 1.64xl04
- 95 l*f-day/kg HM, 38.7% thermal efficiency, 801 capacity factor, 1 year after discharge, near-equilibrium fuel cycle.
- Natural thorium is assumed to contain 100 ppm 23QTh. Discharged thorium is not recycled.
Recovered uranium from fissile fuel is recycled as second-cycle uranium-235
- Initial make-up uranium is 93.1* 23SU.
fuel.
5-5
-------
TABLE 5.3
Actinides in Discharged 255U-"m Second-Cycle IfTGR l-uoj
(1000 the U-Th fueled high-temperature gas-cooled reactori/)
Fissile Fuel
Radionuclide
Thorium^' 228
229
230
231
232
234
Total
Protactinium 233
234
Total
Uranium^' 232
233
234
235
236
237
238
Total
Neptunium 237
Total
Plutonium 236
238
239
240
241
242
Total
Americium 241
242m
243
Total
Curium
242
243
244
245
Total
TOTAL ACTINIDES
kg/yr
4,80xlO"8
5.29xlO"9
f
6.85x10"°
9.39xlO"12
7.47xlO"6
8.42xlO"13
1. 44x10" 5
1.92xlO"12
9.83xlO"18
1.92xlO"12
3.72x10"°
7.35xlO"7
4.64X10"1
2.29
4.37X101
6.90xlO"9
l.lSxlO1
S.SOxlO1
3.79
3.79
1.82xlO"6
2.21
4.75X10"1
2.29X10"1
2.14X10"1
2. 25x10" *
3.35
1.49xlO"2
1.26xlO"4
6. 40x10" 2
7,90xlO"2
8.20xlO"4
_c
5.34x10 3
Z.SOxlO"2
l.lSxlO"4
2.90xlO"2
Ci/yr
3,94x10"*
l.lSxlO"6
-4
1.33x10
S.OOxlO"3
8.17X10'10
1.95xlO"5
a- 3.95xlO"2
_7
6= 5.02x10
3. 98x10" 5
1.95xlO"8
6= 3. 98x10" 5
7.97xlO"2
6.97xlO"6
2.87
_•?
4.91x10
2.77
5.63X10"1
3. 83x10" 3
a= 5.73
6= 5. 63x10" 1
2.67
a= 2.67
9. 68x10" l
3.86xl04
2.91X101
5.19X101
2.41xl04
8.78X10"1
a- 3.87xl04
B- 2.41xl04
4.83X101
1.23
1.19X101
a= 6.02X101
8= 1.23
2.71xl03
2.46
2.33xl03
l.SSxlO"2
a= S.04xl03
6.52xlOJ
a- 4.38x10
B= 2.41X104
Fertile
kg/yr
1.43x10"*
l.SSxlO'4
2.64xlO~2
S.OOxlO"12
6.75xl02
5.09xlO"9
6.75xl02
1.46xlO"4
5. 94x10" 14
1.46xlO"4
1.03xlO"2
l.SOxlO1
3.59
7.32X10"1
1.16X10"1
8.73xlO"13
5.02xlO"6
2.24X101
7.49xlO"3
7.49xlO"3
1.89xlO"13
1.82xlO"3
1.42xlO"4
4.45xlO"5
2.71xlO"5
^ r
1.00x10
2.04xlO"T
1.62xlO"6
5.14xlO"8
4.22xlO"6
S.89xlO"6
3.37xlO"8
1.03xlO"9
8.33xlO"6
4.66xlO'8
8.41xlO"6
6.97xl02
Fuel
Ci/yr
1.17x10*
3. 30x10" 2
5.13X10"1
i.eoxio"3
7. 38x10" 2
1. 18x10" l
-------
TABLE 5.4
Actinides in Discharged 233U-Th Recycled HTGR Fuel
(1000 Mwe U-Th fueled high-temperature gas-cooled reactor— )
Fissile Fuel
I\£tUJ.UI IUC-L J-UC
Thorium^/ 228
229
230
231
232
234
Total
Protactinium 233
234
Total
Uranium^ 232
233
234
235
236
237
238
Total
Neptunium 237
Total
Plutonium 236
238
239
240
241
242
Total
Americium 241
242m
243
Total
Curium 242
243
244
245
Total
TOTAL ACTINIDES
. •
kg/yr
1.30xlO"3
1.82xlO'3
7.43xlO'3
3.96xlO'U
2. 70x10" 5
2. 28x10" 16
1. 06x10" 2
5. 49x10" 12
2.66xlO"21
5.49xlO"12
3.62xlO"2
9.84
2.44X101
9.65
4.92
2. 50x10" 10
S.02xlO'4
4.88X101
5. 95x10" |
5.95X10"1
1.33X10"11
2.66X10"1
2.51xlO"2
l.OlxlO"2
7.75xlO"3
4.04xlO"3
3.13X10"1
S.OOxlO"4
3.23xlO"6
6.72x10'*
l.lSxlO"3
l.SSxlO"5
6. 52x10" 7
2.13x10"*
8. 4 7x10" 7
2.30xlO"4
4.97X101
Ci/yr
1.07xl03
3. 87x10" l
1. 44x10" l
2.11xlO'2
2.95xlO'9
5. 28x10" 9
a- 1.07xl03
B 2.11xlO"2
1.14xlO"4
5.28xlO"12
6= 1.14xlO"4
7.75xl02
9.33X101
l.SlxlO2
2. 07x10" 2
3.12X10"1
2. 04x10" 2
1.67xlO"7
o= 1.02xl03
B= 2. 04x10" 2
4.20X10"1
a= 4. 20x10" -1
7.07xlO"6
4.65xl03
1.54
2.29
8.71xl02
1. 58x10" 2
a= 4.65xl03
6= 8.71xl02
1.62
3. 14x10" 2
1.24X10"1
a- 1.74
-2
B- 3.14x10
5.13X101
S.OOxlO"2
1.77X101
1. 33x10" 4
a= 6.90X101
a= 6.81xl03
B= 8.71xl02
Fertile
kg/yr
S.lSxlO"4
5.58xlO"4
9. 50x10" 2
l.OSxlO"11
2.43xl03
1.83xlO"8
2.43xl03
5.26xlO"4
2.14xlO"13
5.26xlO"4
3.71xlO"2
6.48X101
1.29X101
2.64
4.18X10"1
3. 14x10" 12
l.SlxlO"5
S.OSxlO1
2. 70x10" 2
2. 70x10" Z
6.80xlO"13
6.55xlO"3
S.llxlO"4
1.60xKf4
9. 76x10" 5
3. 60x10" 5
7.35xlO"3
5.83xlO"6
l.SSxlO'7
1.52x10'^
2.12xlO"5
1. 21x10' 7
3.71X10"9
3.00xlO"5
1. 68x10" 7
3.03x10'"
Fuel
Ci/yr
4.23xl02
1.19X10"1
1.85
5.75xlO"3
2.66X10"1
4. 24x10" 1
o= l.SOxlO3
_T
8= 4.30x10
1.09X104
4.24xlO"3
3- 1.09xl04
7.94xl02
6.14xl02
8.00x10*
5.65xlO"3
2. 65x10" 2
2.57xlO"4
6.02xlO"9
a- 1.49xl03
6= 2.57xlO"4
1. 90x10" 2
a= 1. 90x10" 2
3.62xlO"7
l.lSxlO2
3.14xlO"2
3.63xlO"2
l.lOxlO1
1.40xlO"4
a= l.lSxlO2
B= l.lOxlO1
1. 89x10" 2
l.SOxlO"3
2.81xlO"3
a= 2.1 7x10' 2
6= 1. 80x10" 3
4.02X10"1
1.71xlO"4
2.50
2. 63x10" 5
2.90
a= 3.11X100
B= 1.90x10*
a/
— 95 Mw-day/kg HM, 38.7% thermal efficiency, 801 capacity factor, 1 year after discharge,near-equilibrium fuel cycle.
—' Natural thorium is assumed to contain 100 ppm Th. Discharged thorium is not recycled.
—' Uranium in fuel charged is recovered from discharged fertile fuel and previous cycle discharged TJ-Th fissile fuel.
5-7
-------
Radionuclidc-
.»/
TABLE 5.5
Actinides in Discharge Thorium Fuel
Mwe U-Th fueled h
reactor
Half-Life
228
229
230
231
232
234
TOTAL
Protactinium
233
234
TOTAL
Uraniun
TOTAL
Neptunium
TOTAL
232
233
234
235
236
237
238
237
Plutonium^ 236
238
239
240
241
242
TOTAL
Anericiun
TOTAL
Curiun
TOTAL
241
242 !
243
242
243
244
245
1.41xl010yr
24.1 day
27.0 day
6.75 hr
72 yr
1.62xlOS yr
2.47X105 yr
7.1xl08 yr
2.39xl07 yr
6.75 day
4.51x10* yr
2.14xl06 yr
2.85yr
86 yr
24,400 yr
6,580 yr
13.2 yr
3.79xlOS yr
458 yr
152 yr
7950 yr
163 day
32 yr
17.6 yr
9300 yr
emperature gas-cooled HTGR
*3/yr
2.S2.UO*3
1.09x10**
2.64X10"1
1.99X10"10
6.75X103
2.46X10"5
6.7Sxl03
3.6SX10*1
2.87xlO"10
3.65X10"1
1. 39x10"*
1.89xl02
6.20X101
4.9U101
l.OSxlO2
1.69xlO"8
2.88X101
4.32x10*
1.09X101
1.09X101
S.OlxlO*6
5.62
1.18
S.S4X10"1
5.33X10"1
5.40x10-1
8.43
1.92xlO"2
3.01xlO*4
l.SSxlO"1
1.75x10"*
«.90xlO"3
1.30x10"*
6.98xlO"2
2.86x10"*
7.58 x 10"*
Ci/yr
1.91X105
2.52X10"1
S.4O
1.06X10"1
7.38X10"1
S.TOxlO2
a 1.91xl03
B S.TOxlO2
7.S8X106
SiTOxlO"1
B T.SSxlO6
2.98x10*
1.79x10*
3.84xl02
l.OSxlO"1
6.53
1.38
9.60x10**
a S.16X101
0 1.38
7.69
a 7.69
2.66
9.82x10*
7.24XW1
1.25x10?
5.99x10*
2.11
a 9.84x10*
B 5.99x10*
6.2&101
2.93
2.87X101
a 9.09X101
B 2.93
1.62x10*
5.98
5.81x10*
4.49xlO'2
M t •yn-ini
«/
y
c/
i/
95 Mw day/kg f«, 38.71 thermal efficiency, 801 capacity factor
150-day cooling, equilibrium fuel cycle.
Natural thorium is assumed to contain 100 ppm ^h. Discharge
thorium is not recycled.
Includes 59.0 kg/yr of second-cycle uranium from initial make-up
T/, which is not to ho recycled. Com|»sition of discharged
second-cycle uranium: O.St 234U, 3.6t 235I;, 75.5t 236U, 20.11 238u.
I'lutonium is not recycled.
5-8
-------
Each of the three types of uranium in the fresh fuel is formed
into microspheres from 570 to 580 microns in size, which are then
mixed with 820-micron microspheres of thorium and embedded in a
carbonaceous matrix to form a fuel "stick" (Dl). The resulting
fuel sticks are sealed into holes in blocks of high-purity nuclear-
grade graphite, which acts as neutron moderator and structural
support. Heat generated in the fuel sticks is conducted through
the adjacent graphite into helium coolant, which flows through
longitudinal holes penetrating each graphite fuel block.
Each fuel block contains only one of the three types of
uranium-thorium fuel, so that the spatial arrangement throughout
the reactor of blocks containing different types of fissile
uranium provides a means of controlling the spatial distributions
of neutron flux and power density.
The material properties of each of the three fuel types are
given in Table 5.6. The inital and make-up fuel elements, containing
the highly enriched (93.5%) make-up uranium, are formed by 200-
micron microspheres of UC£ and 500-micron microspheres of Th02. The
uranium microspheres are each coated with an inner layer of low-
density pyrolytic carbon to provide voids for fission products and
to act as a buffer layer for fission-product recoil. Surrounding
this is a layer of high-density pyrolytic carbon, a layer of silicon
carbide, and then another layer of high-density pyrolytic carbon
to reduce the diffusional escape of uranium and fission products from
the fuel microspheres. The fuel elements of recycled 235U and
make-up thorium are formed from microspheres similar to those
described above. In the fuel elements containing recycled 233U
and make-up thorium the uranium microspheres are formed from similar
coatings of UC2 particles initially 310 microns in diameter.
The steam generated by the hot helium coolant from the reactor
is at higher temperature and pressure than the steam generated in
water reactors, resulting in an over-all thermal efficiency of
38.7%. For a net electrical output of 1000 Mw the resulting
thermal power is 2583.9 Mw.
The average thermal specific power in the reactor core is cal-
culated to be 65.1 Mw per Mg of uranium and thorium in the fresh
fuel. Each year one fourth of the reactor fuel, contained within
850 graphite fuel blocks (Dl,L1) is discharged and replaced with
unirradiated fuel, so that each fuel element remains within the
reactor for four years. At an average load factor of 80% the
resulting average thermal exposure is 95,000 Mw days per Mg of
uranium and thorium charged (L1,T3).
The coatings surrounding the uranium and thorium fuel particles,
as shown in Table 5.6, not only reduce the escape of fission
products to the gas coolant during reactor operation, but they also
aid the separation of fissile and fertile particles in fuel re-
processing. The reprocessing technique specified for HTGR fuel
involves crushing and burning the graphite blocks in a fluidized
combustor. The ash from the fluidized combustor consists of the
5-9
-------
en
i
Table 5.6
HTGR Fuel Particle Descriptions (PI,L1,T3)
ProjDerty^
Isotope
Kernel Composition
Kernel Diameter (ym)
Type Coating — —'
Coating Thickness (pm)
Buffer Carbon
Inner Dense Carbon
Silicon Carbide (SiC)
OUter Carbon
Total Particle Diameter
(ym)
235U Make-up
Fissile
Particle
-u
uc2
200
TRISO
100
25
25
35
570
Elements-
Fertile
Particle
Th
Th02
500
BISO
85
75
820
233U Recycle
Fissile
Particle
233y
uc2
310
TRISO
50
25
25
35
580
Elements
Fertile
Particle
Th
Th02
500
BISO
85
75
820
235U Recycle
Fissile
Particle
235y
uc2
200
TRISO
100
25
25
35
570
Elements
v^
Fertile
Particle
Th
Th02
500
BISO
85
75
820
a/ For initial and make-up loadings
b/ A TRISO coating consists of a buffer layer surrounding the UC2 kernel, followed by successive
layers of dense pyrolytic carbon, silicon carbide, and dense pyrolytic carbon.
c/ A BISO coating consists of a buffer layer and a single layer of dense pyrolytic carbon.
-------
original IK^ particles still coated with silicon carbide and
oxide particles of UC^-ThC^ from the incineration of the original
ThOo particles coated with pyrolytic carbon. Although the sizes
of these fertile and fissile particles are about the same after
graphite combustion, the thoria particles are about three times
heavier because of the larger diameter of their actinide kernel
and because the SiC and inner carbon coatings of the fissile
particles still remain. The fissile and fertile particles are
separated into two fractions by elutriation with carbon dioxide.
The thoria particles, now containing fission products and bred
uranium, are to be processed by Thorex separation technology,
and the fissile uranium particles containing recoverable uranium,
fission products, and some neptunium and plutonium, are to be
processed by Purex separation technology.
The purpose of making the size separation of the fissile and
fertile particles from each block is to develop a means of con-
trolling the build-up of neutron-absorbing 236U. The fissile
particles used to fabricate each graphite fuel block are one of
three different ty es of uranium described in Table 5.6. Fuel
blocks with different sources of fissile particles are to be proces-
sed separately through graphite combustion and particle clas-
sification, so that the three .different groups of fissile particles
can be collected and treated separately. The particles of uranium
remaining from the first-cycle irradiation of make-up uranium are
to be processed for uranium recycle. The particles of irradiated
bred uranium are to be processed and the recovered uranium is
to be combined with uranium recovered from thorium and recycled.
The uranium particles remaining after the second irradiation cycle
of initial make-up uranium contain a relatively high concentration
of 236U and are to be discarded to transuranic wastes.
Because elutriation does not produce a quantitative separation
there will be some crossover of fertile and fissile particles, and
the crossover will increase as a result of broken particles. The
effects of crossover are to contaminate the recovered 233U with
236U neutron poison and to increase the loss of 233U when fissile
particles are retired. Upper-bound estimates (P4) indicate
that as much as 10% of the fissile particles may cross over
into the fertile stream although less actual crossover is expected.
The calculated effects of crossover on the composition of the
recycled uranium are shown in Table 5.7.
5.2 HTGR Fueled With Thoriumand Denatured Uranium, No
Reprocessing
Until facilities exist for reprocessing uranium-thorium
HTGR fuel, any HTGR must operate on the non-reprocessing cycle,
i.e., it must be fueled with low-enrichment uranium or with thorium
blended with enriched 235U or with plutonium recovered from LWR
discharge fuel. The non-reprocessing fuel cycle of an HTGR fueled
with thorium and enriched 235U is shown in Figure 5.3. In this
case, the isotopic content of 235U in uranium is kept at 20% or
below as a means of reducing the safeguards hazards associated
with 93% 235U in make-up uranium. These safeguards issues are
5-11
-------
1000 Mw
en
i
Thoriym^
3.76 Mg
R
Fobri
i
jel
cation
i
Gas Cooled
Reactor
E = 109 Mw Day/Kg
Fuel Life = 4 yr
rj =0.387
L =o.fln
Fuel
Storage
3.49 Mg Th
2.59 Mg U
0.047 Mg Pu
I9.78%235U
3.23 Mg
Natural
Uranium
0.71 % 236U
136 Mg
Conversion
and Isotope
Separation
Separative Work
132 Mg
0.25% 235U
132 Mg
Fig. 5.3 Annual quantities for the denatured-U-Th-fueled HTGR,
with no recycle (E=fuel capacity, n=overall thermal efficiency,
L=capacity factor)
-------
Table 5.7
Effect of Reprocessing Cross-Over on the Composition of
Recycled Uranium For the HTGR Equilibrium Fuel Cycle
No cross-over With cross-over--
charge,% discharge,% charge,% discharge,:
Recycled 235U
232U
233U
Particles
235U
236y
238U
1
0
0
.22
29.51
52.24
17.03
0
0
0.81
3.96
75.28
19.95
0.002
4.17
1.99
28.15
49.53
16.16
0.001
0.39
.74
,08
74.17
19.61
1
4.
Recycled Bred Uranium
232U
233U
23"U
235U
236U
0.
52.
22.
6.
04
31
04
73
238
u
18.87
0
0.04
7.83
31.65
12.27
48.21
0
0.04
42.20
17.88
6.32
30.61
2.95
0.03
5.24
21.27
8.39
60.33
4.75
a/ Assumed reprocessing cross-over: 10% of the fissile particles into
the fertile stream, 5% of the fertile particles into the fissile stream.
5-13
-------
discussed in more detail in Chapter 7. Calculated material quan-
tities for the near-equilibrium fuel cycle of the HTGR operating
with denatured-uranium-thorium fuel, normalized to data of Haffner,
et. al. (H3), are given in Tables 5.8 and 5.9.
5.3 Resource Utilization by Current and Modified HTGR Designs
The thirty-year lifetime ore requirements for the HTGR fueled
with denatured-U-Th, with no recycle as in Fig. 5.3, and for the
reference design of the HTGR, operating according to the overall
flowsheet of Figure 5.1, are shown in Table 5.10. The non-repro-
cessing HTGR cycle requires 50% more uranium ore than does the U-Th
reprocessing cycle of the HTGR reference design, but it requires
21% less uranium ore than does the uranium-fueled non-reprocessing cycle
of the PWR (Fig. 3.1).
Also shown in Table 5.10 is an estimate for a similar HTGR
which uses plutonium recovered from LWR discharge fuel as make-up
fissile material (P2). Both uranium and plutonium in this version
of the HTGR are assumed to be recycled. The ore required for this
case includes ore for a uranium-fueled LWR to supply the make-up
Plutonium, with the total reactor system scaled to an electrical
energy generation of 30x0.8 Gw yr. Much of the ore required .for
these recycle cases is that necessary to supply the fissile material
for the initial loading and for extra fissile make-up during the many
irradiation cycles before the near-equilibrium.fuel cycles are
reached. However, a second-generation HTGR could start-up with
the fuel cycle inventory left from a retired first-generation HTGR.
The lower lifetime ore requirements of such second-generation HTGR's
are shown in Table 5.10. The HTGR can also be operated on low-
enrichment uranium, with or without U-Pu recycle.
Comparing the data in Tables 3.1, 4.1 and 5.10 for thorium
fuel cycles with235U make-up, it is apparent that the current HTGR
reference design is intermediate in ore requirement between the less
efficient thorium version of the LWR and the more efficient thorium
versions of the CANDU reactor.
5-14
-------
TABLE 5.8
Actinides in the Fuel Charged to the Denatured HTGR
(1000 MWe, Stowaway cycle^)
Radionuclide
Thorium 232
total
Uranium 235
238
total
kg/yr
3.72 x 103
3.72 x 103
6.33 x 102
2.57 x 103
3.20 x 103
Ci/yr^
4.07 x ID'1
a = 4.07 x ID'1
1.36
8.55 x ID'1
a = 2.22
weight %
100.00
100.00
19.78
80.22
100.00
a/ 109 Mw day/kg U+Th, 38.7% thermal efficiency, 80% capacity factor,
near-equilibrium fuel cycle
b/ "a" denotes radioactive decay by alpha emission
5-15
-------
TABLE 5.9
Actinides In The Fuel Discharged From The Denatured HTGR
(1000 MWe, Stowaway cycle -/)
§/
Radionuclide
Thorium-'
total
c/
Uranium-'
total
Plutonium-'
total
232
233
234
235
236
238
239
240
241
242
kg/yr
3.49 x 103
3.49 x 103
8.50 x 101
1-60 x 101
5.50 x 101
8.70 x 101
2.35 x 103
2.59 x 103
1.70 x 101
1.00 x 101
8.00
1.20 x 101
4.70 x 101
Ci/yr
3.81 x TO'1
a = 3.81 x ID'1
8.06 x 102
9.90 x 101
1.18 x ID'1
5.52
7.82 x 10-1
a = 9.11 x 102
1.04 x 103
2.27 x 103
9.00 x 10s
4.68 x 101
a = 3.36 x 103
B = 9.00 x 105
weight %
100.00
100.00
3.28
0.62
2.12
3.36
90.62
100.00
36.17
21.28
17.02
25.53
100.00
a/ 109 Mw day/kg U+Th, 38.7% thermal efficiency, 80% capacity factor,
near-equilibrium fuel cycle. Np, Am, Cm are not included.
b/ Other thorium isotopes are not included.
c/ 23ZU and 237U are not included.
d_/ 238Pu is not included.
e/ Activities from alpha decay and beta decay are indicated by "a"
and "8", respectively.
5-16
-------
TABLE 5.10
30-Year Lifetime Ore Requirements for High-Temperature Gas-Cooled Reactors
(Current HT6R Reference Design, 1000 Mw Electrical Power, 80% Capacity Factor)
Natural Uranium,
Metric Tons (Mg)
Natural Uranium Thorium,
relative to metric tons
Fuel
Cycle
(a)Denatured-U-Th fuel,
no recycle
(b)235U-Th fuel,
U recycle
(c)Pu-Th fuel, ..
U-Pu recycle^'
(d)2nd gen. 235U-
Th, U recycle
(e)2nd gen. Pu-Th,
U-Pu recycle?/
Conversion
Ratio
0.50
0.66
0.64
0.66
0.64
0.20%
depleted U
3870
2290
3850
1920
1974
0.25%
depleted U
4270
2840
4130
2130
3000
U-fueled PWR
w/o recycle3'
0.79
0.53
0.77
0.40
0.56
151
247
79.9
221
70.3
*/ 30.4 Mw-day/kg HM, 34.2% thermal efficiency, 80% capacity factor, near-equilibrium
cycle.
-/ 23l4U is not included.
-' 150 days cooling of discharged fuel before reprocessing. 1.5% loss in
reprocessing, 1.5% loss in fabrication. Natural uranium is added to the
recycled plutonium to dilute the recycle fuel to proper enrichment.
d/ Includes U-fueled PWR to supply make-up Pu. Total system operates for
30 yr at 1000 Mw.
-^ Starts with equilibrium reactor and fuel-cycle fissile inventory.
5-17
-------
The current reference HTGR, which is the basis of the flowsheet of
Figure 5.1 and the data in Table 5.10, is a design optimized for current
or near-term fuel-cycle cost parameters, including uranium ore prices.
However, a feature of the HTGR fuel concept is the flexibility for making
relatively large changes in the fuel loading without altering the fuel
thermal performance or the overall mechanical design, or with only modest
changes in these design parameters. For a future era of higher uranium
ore prices and greater incentive to improve ore utilization, the conversion
ratio of the HTGR can be increased by the following modifications (B5):
1. Increase the thorium loading in the core, which increases
neutron absorption in thorium relative to non-productive
absorption and leakage.
2. Decrease the core power density. For the same fissile con-
centration, this decreases the neutron flux and reduces the
flux-dependent neutron absorption in 135Xe and 233Pa. The
greater core volume, for the same thermal power, provides volume
for further increases in the thorium loading.
3. Decrease the time interval between refueling, thereby decreasing
the loss of neutrons to control absorbers.
4. Reduce the thickness of the coatings on the fuel particles,
allowing greater theorium loading in the graphite-prism fuel
holes and thereby allowing greater thorium loading per prism.
5. Distribute the fuel particles uniformly throughout the graphite
prism.
Estimated improvements in the conversion ratio and ore requirements
made possible by such approaches are shown in Table 5.11.
The first two modifications, which increase the conversion ratio to
0.76, can be achieved with the current fuel element design. This reduces
the life-time ore requirement to a level about 58% greater than that of
the CANDU reactor operating on the same fuel cycle.
The most significant parameter in increasing the HTGR conversion
fatio is the increased thorium loading. This requires corresponding
increases in the initial and start-up loadings of fissile 235U, resulting
in a greater investment in fuel early in the reactor life. The higher
initial investment contributes to a higher levelized fuel cycle cost, but
if the price of uranium increases more rapidly than does the effective
discount factor during the plant life, the lower annual ore requirements
for fuel reloads throughout the plant life could possibly offset this
higher initial fuel investment.
One feature of the HTGR which benefits its fuel cycle cost and its
resource utilization is the very high fuel exposure of 94 Mw day per
kilogram of heavy metal. This means that for fuel reprocessing and re-
fabrication to make the same contribution to the cost of electrical
5-18
-------
TABLE 5.11
Conversion Ratio Improvements Possible for the HTGR Fueled with 233U,
Th, and Recycled Uranium
Lifetime ore requirement-
Modification^
Reference HTGR, 235U-Th fuel,
U recycle
Increases thorium loading by 25%
Change from yearly fueling to
semiannual fueling
Reduce core average power density
from 8.4 to 6.0 watts/cm3
Use modified fuel elements
and/joi* improved fuel particles
Conversion ratio^
0.66
0.71
0.76
0.82
0.90
0.95
relative to
reference HTGR
1.0
0.89
0.80
0.68
0.57
relative to U- fueled
PWR w/o recycle^/
0.53
0.47
0.42
0.36
0.30
a/ From Brogli, et al., (B4).
b/ Includes initial loading and reloads over 30-yr lifetime, calculated from data
of Brogli, et al., (B4).
-/ 30.4 Mw-day/kg HM, 34.2% thermal efficiency, 80% capacity factor, near-equilibrium
cycle.
5-19
-------
energy, the unit costs of these operations, expressed in cost per unit
amount of heavy metal processed, can be correspondingly greater for these
operations in the HTGR fuel cycle. However, whether these HTGR operations
can be carried out within the greater allowable unit costs is uncertain
at this time. The HTGR reprocessing operations are yet to be carried out
on a pilot-plant scale, so the technological foundation for estimating
the cost of commerical-scale operations is now quite limited.
The burnup advantage of HTGR fuel over LWR and CANDU fuels decreases,
but does not disappear, as modifications are made to improve the HTGR con-
version ratio. The improvements in conversion ratio and ore requirements
listed in Table 5.10 were calculated on the assumption that the fuel would
be irradiated for a constant time interval of four years, as in the present
HTGR reference design. Therefore, as the thorium loading and fissile loading
are increased to improve the conversion ratio, the burnup correspondingly
decreases.
The HTGR reactor design is well founded and is readily adaptable to
the modifications described herein. However, the technical complexities
and lack of engineering-scale experience in the HTGR fuel cycle suggest
caution in economic comparisons with other fuel cycles. Thorough and
periodic engineering evaluation of the economics of the HTGR fuel cycle and
of alternative thorium-based fuel cycles is important. Similar reactors
are under development in Germany, where designs of the prismatic type as
well as advanced pebble bed designs are being considered.
5-20
-------
6. Fuel Cycles for Fast-Breeder Reactors*
6.1 The Reference PuO^-UO,, LMFBR
Consideration of fast-breeder fuel cycles is relevant to the
issues of thorium fueling because:
(a) The possible resource need for a thorium fuel cycle al-
ternative in near-term reactors to reduce the consumption"of uranium
ore disappears if and when the uranium-piutonium breeders are introduced.
However, if the uranium-pi utonium breeder is deferred or delayed,
an alternative thorium fuel cycle in LWR, HTGR, or CANDU reactors may
become an important means of conserving uranium resources.
(b) Fast breeders fueled with metallic thorium and recycled
uranium can achieve higher breeding ratios and lower sodium-void
coefficients of reactivity than PuO?-UO? fast breeders.
(c) Fast breeders with U-Pu cores and blanketed in part with
thorium can consume the plutonium produced by dispersed national
reactors fueled with denatured uranium and thorium and can supply the
make-up 233U to fuel these safeguarded national reactors. This is
considered in more detail in Chapter 7.
The fast-breeder programs in this country and elsewhere are
focused on the development of the sodium-cooled breeder reactor fueled
with PuO~ and UO^. In a given type of fuel material, such as oxide
fuel, higher breeding ratios and shorter doubling times are possible
with the uranium-plutonium fuel to start up the first generation
breeders. Moreover, the Purex reprocessing technology is available
for the uranium-pl utonium cycle. Figure 6.1 shows a flowsheet for
a possible early LMFBR operating on an equilibrium fuel cycle (Gl)
fueled with natural or depleted uranium. The excess plutoniunvpVoduc-
tion from this breeder can be used to start up subsequent breeders,
provided that the doubling time for increasing breeder capacity is
no shorter than the doubling time for excess fissile production by
the breeder.
A large amount of depleted uranium from isotope separation will
have been stockpiled by the time when present low-cost uranium re-
sources are consumed by water reactors. Assuming that breeders replace
the water reactors then being retired in the next century, and assuming
that the total fission power continues at a constant level, the
stockpiled depleted uranium is an already-mined resource sufficient to
fuel these breeders for thousands of years. The fast breeder is the
most efficient of all fission systems in terms of long-term ore util-
ization.
The only ore requirement attributable to the breeder is that
associated with the production of plutonium for start-up loadings of
the first-gene ration breeders. This plutonium must be obtained
*Much of the text of this section was adapted by the first author for
incorporation in the APS report on nuclear fuel cycles and waste manage-
ment (HI).
6-1
-------
Uranium
0.71 to 0.25% 235U
1.19 Mg
1000 Mw
Pigford, 1977
Blanket
Fuel
Fabrication
en
ro
Uranium
Recycle
14.7 Mg
Blanket
Fast Breeder Core
E=68.3Mw day/kg
Fuel Life =2.08 yr
0=0.35
L = 0.80
Fuel
Reprocessing
Core
Fuel
Fabrication
Uranium
9.l7Mg
Plutonium
2.03 Mg
Fission
Products
0.876Mg
Plutonium Product
80% Fissile
0.316 Mg
Fig. 6.1 Annual quantities for LMFBR fueled with natural or
depleted uranium (equilibrium fuel cycle, E=fuel exposure,
n=thermal energy, Incapacity factor)
-------
from light-water reactors, and these reactors will then require
more ore because they are thereby deprived of the benefits of
Plutonium recycle. When operating without Pu recycle, the 1-Gw
LWR produces 171.4 kg/yr of fissile Pu. LWR's must operate for
43.8 Qw-yr without Pu recycle in order to produce the 7500 kg of
fissile Pu required to start-up a 1-Gw LMFBR. The uranium ore
attributable to Pu production is the difference between the ore
required for operating LWR's with U recycle only and that required
with U-Pu recycle. Using the data of cases (b) and (c) of Table
3.3 and scaling to 43.8 Qw-yr, we estimate 1180 Mg of natural uranium,
for 0.25% depleted uranium tails attributable to Pu start-up as shown
in Table 6.1.
If the doubling time for subsequent growth in breeder capacity
is no greater than the doubling time for the breeder to produce
excess plutonium, no ore is required for subsequent breeder genera-
tions. For each gigawatt of first-generation breeder capacity in-
stalled in the 1990's, 43.8 Gw-yr of light water reactors must be
operated without plutonium recycle during the 1980's and early
1990's to furnish the start-up plutonium. Therefore, breeder intro-
duction in the 1990's would require the existence of industrial-
scale LWR reprocessing several years before that time. The present
schedule is uncertain.
The data in Table 6.1 indicate that over a 30-year operating
life, three uranium-fueled light water reactors could produce
enough plutonium to start up two fast breeders, if no plutonium
were to be recycled in water reactors. Alternatively, nine water
reactors operating during their last ten years of life without
plutonium recycle will generate enough plutonium to eventually
start up two breeders. The 1974 ERDA projections of U.S. nuclear
power growth indicated a growth to 124 GW of fast breeder capacity
by the end of the century! along with 644 GW of light water reactors.
Calculations (P2) of the amount of start-up plutonium required for
such a large scale of breeder introduction snowed that plutonium
recycle in water reactors would have to be discontinued in the early
1990's to insure sufficient plutonium for breeder start-up. However,
events since 1974 suggest that such a rapid introduction of breeders
is not likely, and delays in LWR fuel reprocessing and in the con-
struction of additional LWR fuel reprocessing facilities seem more
likely to result in an over supply in the 1990's of plutonium which
can be extracted from water reactor fuel.
From the above it is apparent that there are several situations,
any one of which could warrant operating water reactors entirely with
uranium fueling so that all of the plutonium produced would be
available for breeder start up. Examples are:
6-3
-------
TABLE 6.1. Fissile, Ore, and Enrichment Requirements to Start a
First-Generation Fast Breeder Reactor with Water-Reactor Pluto-
nium (1000 Mw electrical power, 80% capacity factor).
Fissile Pu required for 7500 kg
fast breeder start-upi/
Operation of U-Fueled
water reactor to generate 43.8 Gw Yr
Pu start-up inventory
U in ore attributable to production
of startup Pu
0.20% depleted U 1060 Mg natural U
0.25% depleted U 1180 Mg natural U
Additional separative work due
to loss of Pu-recycle in
water reactors:
0.20% depleted uranium 1200 Mg
0.25% depleted uranium 1020 Mg
Example: To start up 1 GW of FBR requires that 4.38 Gw of LWR be
operated for 10 yr. without Pu recycle. Total natural U used =
6540 Mg, assuming 0.25% depleted U. Total natural uranium attributable
to breeder start-up = 1180 Mg.
-Based upon 3000 kg fissile Pu for the initial core plus 4500 kg
for replacement loadings before Pu in discharge fuel is recycled (Gl).
6-4
-------
(a) a very limited supply of uranium ore
(b) a sufficiently large ratio of first-generation breeders to
previously installed water reactors
(c) a desire to move as rapidly as possible into a breeder power
system.
6.2 Fast Breeder Start-up with 235U
Plutonium is the best of all the fissile isotopes in achieving
high breeding ratios and low doubling times in fast breeders (B12,
61, P9, Yl, Y2}. Although enriched (20 to 22%) 235U from isotope
separation could be used for breeder start-up, the relative penalties
associated with 235U results in larger fissile inventory and a lower
breeding ratio than with plutonium fueling.
Calculated amounts of natural uranium and separative work
for 235U start-up are presented in Table 6.2. It is shown that for
a commercial fast breeder optimized for an equilibrium piutoniurn-uranium
fuel cycle, the amount of fissile uranium required for start-up is from
1.5 to 2.4 times as large as the amount of fissile plutonium that would
be required, depending upon the method of reprocessing the core fuel.
235U start-up would consume greater quantities of uranium ore than that
attributable to Pu start-up from LWR's, and also would require greater
quantities of electrical energy for isotope separation. The corres-
ponding total cost of the fissile material for start-up would be greater
by a factor of 2.3 to 3.7 for enriched uranium than for plutonium (HI,
P9).
The breeding ratio is significantly lower during start-up cycles
with 235U, and this effect persists for many subsequent reloads until
most of the 235U has been recycled and consumed. The new deficit in
breeding-gain production of fissile plutonium due to 235U start-up of
a 1000 Mw LMFBR is about 1700 kg. This considerably increases the
breeding-gain doubling time and will delay the start-up of second-
generation breeders, assuming these are to be fueled initially with Pu
from first-generation breeders.
6.3 Summary of Resource Requirements for the Reference LMFBR
The total lifetime ore requirements for the reference LMFBR,
including the ore for start-up and the ore for life-time refueling,
are shown in Table 6.3. For the first-gene ration breeders, which
require start-up fissile material from an external non-breeder source,
the lifetime ore requirement is still less than any of the first-
generation light-water-reactor cases listed in Table 3.3 and is
less than any of the first-generation HTGR cases listed in Table
5.5, when all cases are calculated with the same concentration of
235U in depleted uranium from isotope separation. However, any one of
the first-generation CAMDU cases with recycle (Table 4.1} requires
less lifetime ore than does the LMFBR with 235U start-up but requires
more lifetime ore than does the first-generation LMFBR with plutonium
start-up. The second-generation LMFBR, which received its start-up
fissile inventory from first-generation breeders, requires no ore
if it is fueled with depleted uranium which has been previously stock-
piled from isotope separation. If fueled with natural uranium, the
6-5
-------
TABLE 6.2. Fissile, Ore, and Enrichment Requirements to Start
A First-Generation Fast Breeder Reactor on Enriched Uranium
(1000 Mw electrical power, 80% capacity factor).
A. Uranium in discharge core fuel is reproduced separately from
uranium in axial and radial blankets
Fissile 235U required for fast breeder start-up£( Mg 11.25
Natural uranium in ore required for 20% 235U, Mg
0.20% depleted uranium 2160
0.25% depleted uranium 2400
Separative work required for 20% 235U, Mg
0.20% depleted uranium 256°
0.25% depleted uranium 2330
B. Fuel elements containing core and axial blanket are chopped
and processed without core-blanket separation, so enriched
uranium is not recycled
Fissile 235U required for fast breeder start-up^, Mg 18.00
Natural uranium in ore required for 20% 235U, Mg
0.20% depleted uranium 2480
0.25% depleted uranium 3840
Separative work required for 20% 235U, Mg
0.20% depleted uranium 4100
0.25% depleted uranium 3730
-Based upon 4.5 Mg 235U for the initial core plus sufficient re-
placement loadings until reactor is self-sustaining on recycle fissile
material. Although lower 235U loadings are possible for a breeder core
optimized for 235U fueling, the purpose here is to start-up a core op-
timized for steady-state fueling on bred plutonium (Gl, P9).
6-6
-------
TABLE 6.3. 30-Year Lifetime Ore and Enrichment Requirements
for Fast-Breeder Reactors!/(1000 Mw electrical power,
capacity factor)
Source of fissile material
for start-up
Amount of start-up
fissile material, Mg
Natural uranium in ore to
produce start-up
fissile inventory, Mg
Separative work attributable
to breeder start-up, Mg
Natural uranium in ore for
inventory and lifetime
refueling
if fueled with natural
uranium, Mg
if fueled with depleted
uranium from stockpile, Mg
Total natural uranium in ore
for start-up and for lifetime
refueling
if fueled with natural
uranium,
Mg
relative to lifetime
requirement for U- fueled
PWR
if fueled with depleted
uranium from stockpile,
Mg
relative to lifetime
requirement for U-fueled
PWR
First-generation
breeder
Pu from U-fueled
LWR
7.5
1180
1020
69.3 &
0
1250
0.226
1180
0.220
20% enriched
u k/
11.25
2400
2330
69.3^
0
2470
0.454
2400
0.447
Second-generation
breeder
Pu from first-
generation
breeder
7.5
0
0
35.7
0
35.7
0.007
0
0
-^Calculated for 0.25% 235U in depleted uranium from isotope separation.
— Increase material quantities by 60% if core and axial blanket are chopped and
processed without core-blanket separation, so enriched uranium is not recycled (P9).
- Includes start-up inventory of uranium in reactor and fuel cycle. Assume two-
year hold-up in external fuel cycle.
6-7
-------
u °f the seco"d-generation LMFBR is less
than l/o of that of the best second-generation, i e , equilibrium
inventory,cases of the PWR of Chapter 3; it is 1.7% of the life-
time ore requirement of a second-generation reference-design HTGR
of Chapter 5; it is about 4% of the lifetime ore requirement for
the best second-generation CANDU case of Chapter 4.
Any other breeder, such as thermal breeders that may result from
modification of the thermal reactors discussed herein, can ultimately
operate on an equilibrium fuel cycle that requires no greater ore
for refueling than does the second-generation LMFBR. However, the
time to reach equilibrium for these thermal breeders is greater than
for the LMFBR, and there will be a greater ore consumption by the
thermal breeders before breeding equilibrium is reached.
6.4 Thorium Fuel Cycles For Fast Breeder Reactors
There is some interest in breeders fueled with thorium and re-
cycled uranium. For example, if the breeder program were to be
significantly delayed and if thorium fueling of thermal reactors were
to be introduced, as discussed earlier, to conserve uranium re-
sources, these thermal reactors would eventually become sources of
233U instead of plutonium for breeder start-up. Although 233U is
far better than 235U for this purpose and results in reasonable
breeding ratio, it is still inferior to plutonium.
When Th02 is substituted for UOp in the core fuel, case (b), the
breeding ratio decreases. This is a result of the lower fast fission
cross section of 232Th and also from the partial replacement of 239Pu
by 233U as the latter builds up and fissions during the irradiation-
cycle. Substituting Th02 for U02 in the blanket only slightly decreases
the breeding ratio because of the relatively few fissions in the
blanket. A core fueled with 233U02 - ThOp results in an even lower
breeding ratio.
Since the irradiation behavior of DCL - ThOp fuel appears to be
similar to that of PuO- - UCL fuel, it is likely that LMFBR's designed
and introduced with Pu62 - U02 fueling could be converted later to
U02 - Th02 fueling. A longer doubling time would result, but the
extent to which this would be a problem would depend upon the desired
rate of breeder introduction.
Uraniurn-thorium fueling in breeder cores may have some safety
advantage because of the smaller increase in reactivity from sodium
voiding than with plutonium-uraniurn fueling. However, there are
other means of reducing the reactivity effects of sodium voiding,
if this proves to be necessary in the LMFBR development program.
Introducing thorium fueling in breeders would introduce many of the
problems that would be encountered with thorium fueling in thermal
reactors. The build-up of 232U in the irradiated fuel and the high-
energy gammas of the 232U-decay daughters would require more shielding
and remote handling in fabricating recycled fuel, and it complicates
fuel reprocessing. The 232U build-up in a thorium-fueled fast reactor
is likely to be considerably greater than in thorium-fueled thermal
reactors. Also, the reprocessing would have to be based upon Thorex
6-8
-------
technology, which is not as well developed as Purex reprocessing
and is expected to be more difficult and expensive. The control
of shut-down reactivity is more difficult with 233U fuel because
of the relatively long (27.0 day) half life of 233Pa, the 233U
precursor. The long half life results in increased precursor
concentrations during operation. Significant reactivity is added
by 233Pa decay after reactor shutdown, and more control absorbers
are required with uranium-thorium fueling. Also, the delayed neutron
fraction for 233U is lower than that for 239Pu, so lower worth for
individual control absorbers and slower withdrawal rates to avoid
prompt criticality may be required. These operational problems
can all be accommodated through proper design, but they can affect
the economics of uranium-thorium fueling.
More advanced sodium-cooled breeders designed for higher
breeding ratios and higher specific powers may be based upon fuel
materials in the form of carbides, nitrides or metals. As shown
in Table 6.4, these advanced fuel materials offer better theoretical
thermal and neutronic performance, but less is known about their
irradiation behavior than is known about oxide fuels. Also 1IfC
formation in nitride fuels may result in greater expense in envir-
onmental controls and in waste management. Although uranium-
metal fuels have been considered unacceptable for the high burnups
required for breeder cores, experience of the EBR-II experimental
breeder now indicates that alloyed uranium metal may be suitable.
Fuels of thorium-base alloy may be an even more attractive possibility.
The isotropic face-centered-cubic structure of thorium metal is more
stable than uranium to irradiation damage and swelling (S2). Thorium
undergoes its solid-phase transformation at 1365°C, which is much
higher than the 660°C transformation temperature of uranium metal.
Also thorium melts at 1725°C, as compared with 1132°C for uranium.
However, because of the limited solubility of uranium and plutonium
in thorium, the irradiation behavior of U-Th and U-Pu-Th alloys for
core fuel may not be as good as that expected for thorium metal.
The irradiation behavior of such alloys at operating temperatures and
design burnups is not sufficiently known.
The higher thermal conductivity of thorium-based alloys could
result in higher specific power than with oxide fuel. Also, the
higher atomic density of the metal should result in a breeding ratio
higher than that attainable with oxides, as shown by cases (d) and (e)
in Table 6.4. The higher specific power and breeding ratio both result
in a lower doubling time for the thorium-alloy fuel. Also, with metal
fuel the reactivity effects from sodium voiding are further reduced
below those predicted for the oxides. These possible advantages from
thorium-alloy fuel in breeders, as compared to thorium oxide fuel,
must be weighed against the greater uncertainties in irradiation behavior
and possibly more expensive fuel fabrication. Also, thorium alloy
6-9
-------
TABLE 6.4. Comparison of Pu-U and U-Th Fueling in LMFBR's -
(1000 Mwe, 0.8 load factor).
a
Core Fuel
Material
Mixed-Oxi
) 239Pu02-U02
b) 239Pu00-ThO
Blanket
Material
de Fuels
uo2
o UOo-ThOo-/
Breeding
Ratio
Excess Fissile
kg/Gw yr
233jj 235y 239pu
1.23 — 31.3 165.1
1.15 43.3 — -34.6
Production
2ltlPu net
7.6 141.4
3.8 96.5
Metallic Fuels
c) 239Pu233U-Th
d) 233U-Th^/ U-Th
U-Th-/
1.31 335.8 — -104.9 1.7 232.6
1.21 -44.1 210.1 166.0
-Calculated for equilibrium fuel cycle (52).
-/Depleted U09 radial blanket, Th00 axial blanket.
c/
— Metal core, depleted U metal radial blanket, Th metal axial blanket.
6-10
-------
fuels will be subject to the same problems of thorium technology
described above. Therefore, the present state of knowledge on
thorium fueling in fast breeders does not suggest diversion from
the PuO,, - U02 fuel now under development. Advanced carbide and
thorium-alloy fuels do offer promise for longer-range improvements
in advanced breeder designs.
Thorium-alloy fuels for breeder cores are not adaptable to the
concept of a breeder fueled with denatured uranium for safeguards
fuel cycles. If the recycled 233U were diluted by natural uranium
to a fissile content of 12 to 20% (see Chapter 7), as is proposed
for denatured uranium fuel cycles, core reactivity limitations do
not allow further dilution with thorium. Also, even for an all-
uranium 233U- 238y denatured core, it will be difficult to reach
criticality if the 233U content must be kept as low as 12%, as suggested
in Chapter 7.
Helium-cooled fast breeders have also been studied and are still
receiving research and development support. Higher breeding ratios
are theoretically possible. However, less is known about the struc-
tural stability of the fuel and the irradiation behavior of fuel
cladding. Also, approaches to emergency cooling which differ from
those designed for the LMFBR are necessary.
6-11
-------
7. Technical Safeguards Features of Thorium Fuel Cycles and
Denatured Fuel Cycles
7.1 Safeguards in Normal Thorium Fueling
"Normal" thorium fueling consists of thorium mixed with highly
enriched fissile make-up and recycled uranium, as has'been illustrated
in Sections 3, 4 and 5. For the equilibrium fuel cycles of a uranium-
thorium fueled light water reactor, the recycled uranium typically
contains about 55% 233U and 10% 235u, which is a fissile content
sufficient for nuclear explosives. However, the recycled uranium will
contain appreciable concentrations of 23*U. As illustrated in Figure
2.3, the gamma activity and external gamma dose rate due to 232U daughters
grows rapidly after fuel reprocessing. After 100 days a metallic uranium
part as small as one kilogram and containing 100 ppm 232U will produce a
gamma dose rate as large as 0.1 rem/hr at one meter. Recycled uranium in
a uranium-thorium cycle may contain several hundred to several thousand
ppm of 232U, depending upon the 230Th content of the make-up thorium and
upon the fuel lattice, so the surface dose rate will be considerably
greater than shown in Figure 2.3. Therefore, recycled uranium from
thorium irradiation will require more shielding than reactor-grade
plutonium. This could affect the practicality of using 233U-rich recycled
uranium for explosives.
The fissile make-up for normal thorium fueling, as illustrated in Sec-
tions 3, 4, 5, consists either of highly enriched (93%) 235U or plutonium
recovered from discharge fuel from uranium-fueled water reactors. Highly
enriched 235U is the least radioactive of all the separated fissile mate-
rials. It can be handled with relatively little hazard from its radio-
activity. Although its fast-assembly critical mass is greater than that of
plutonium, 235U has a relatively low neutron background from spontaneous
fission and from (a,n) reactions. It can be assembled into simple gun-type
devices. Uranium metal is less reactive chemically than plutonium metal.
Therefore, the use of highly enriched 235U introduces what may be the most
significant of all the safeguards concerns in the various nuclear fuel cycles.
If thorium fuels are reprocessed soon after discharge from the reactor,
appreciable quantities of undecayed 2;i3Pa may be present. A relatively
simple chemical separation could yield pure /;33Pa. Its subsequent decay
to 2J3U would yield a pure fissile material for explosives.
7.2 Low-Enrichment Denatured-Uranium Fuel Cycles
Various fuel cycles have been suggested as means of restricting the
possibilities of diverting fissionable material from nuclear power fuel
cycles. The non-reprocessing fuel cycle for a uranium-fueled light water
reactor, as illustrated in Figure 3.1, could be adapted to an international
safeguard fuel cycle. The low-enrichment uranium fuel, containing about 3%
23bU, is "denatured" in the sense that additional isotopic enrichment
would be required for it to be used as material for a nuclear explosive.
The discharge fuel, which contains significant quantities of fissionable
plutonium, could be stored under international inspection or control (i.e.,
an international stowaway cycle). This cycle will ultimately entail
7-1
-------
higher costs, since it is the greatest consumer of natural uranium
and requires a relatively large supply of slightly enriched uranium.
The alternative of reprocessing the discharge fuel and storing the
recovered plutonium under international inspection or control may
impose additional safeguards and financial burdens. The stored
plutonium must be protected, and the cost of storing separated
plutonium is high compared with the cost of storing discharge fuel
(H1,P3). Nevertheless, the stowaway cycle is technically the simplest
of the alternatives discussed herein and can be consistent with their
later implementation. If such international safeguards fuel cycles
are to be utilized, the stowaway version represents a possible first
step that could be implemented with existing technology.
Another alternative is to fuel all such national reactors, to be
under individual safeguard control, with slightly enriched uranium
and to ship the discharge fuel to a centralized fuel reprocessing
center under international control. The recovered plutonium would be
consumed on-site in piutoniurn-burner reactors. The electrical distri-
bution system receiving the energy generated by pi utoniurn-burner
reactors would require relatively little uranium ore. The uranium
ore thus saved could then be used as feed to a centralized uranium-
enrichment plant to supply the slightly enriched uranium fuel for the
externally located uranium-fueled reactors. The total uranium ore
consumption for the entire generating system would be the same as if
all reactors were nationalized and operating with self-generated
uranium-plutonium recycle. However, financial and uranium exchanges
between participating countries are required. An overall flowsheet
of this safeguards fuel cycle at equilibrium is shown in Figure 6.1.
Since the fuel discharged from the uranium-fueled reactors would still
contain plutonium, the storage and shipment of the discharge fuel
would have to be under safeguard control. Again, this cycle represents
a step based on an existing technology and could be implemented in the
near future.
In calculating the actinide quantities for the national-international
fuel cycles shown herein, it has been assumed that 1% of the actinides
are lost to reprocessing wastes and 1% to fabrication wastes. It is
obviously necessary that the fissile content in these wastes be identified
and safeguarded.
7.3 Denatured-Uraniurn-Thorium Cycles with Pressurized-Water Reactors
An alternative to the uranium cycles is the thorium-uranium cycle,
in which 233U is formed by neutron absorption in 232Th. The fissile
uranium in the fuel is to be denatured by dilution with natural or
depleted uranium. The isotopic concentrations at which the fast-assembly
critical masses for *35u_238u and 233u_238u mixtures become very large,
and presumably unsuitable for explosives, are (HI, P5):
235[j
y— = 0.20
233 238
U + U
= 0.12
7-2
-------
For fuels containing masses M233, M235, and M238 of 233U, 235U
and 23bU, respectively, the required dilution by 238U is assumed to
be obtained by the linear combination:
M > / 1 1 \ M J_ / 1
\?
02
Thorium is then added as additional fertile material so that the overall
fissile concentration in the fuel is a few percent, typical of fuel for
light-water reactors. This fresh fuel of denatured uranium and thorium
is similar to low enrichment (i.e., "denatured") uranium fuel in that
isotopic enrichment would be necessary in either case to produce uranium
suitable for an explosive assembly. It differs in that much of the 238U
has been replaced by thorium,, so that the production of chemically
separable plutonium has been suppressed. However, appreciable quantities
of plutonium are still present in the spent fuel, and the same set of
issues as to its disposition still arise.
In Figure 7.2 is shown the overall equilibrium flowsheet for the
international fuel cycle in which pressurized-water national reactors
are fueled with denatured uranium and thorium, and plutonium is consumed
in international piutonium-uraniurn-fueled pressurized-water reactors.
The model reactor used in these calculations is that described in Section
3. The plutonium production per unit amount of 238U in the national
reactor fueled with uranium-thorium is 1.85 times greater than in uranium-
fueled reactor of Figure 3.1. The lower concentration of Z38U in uranium-
thorium fuel decreases the self shielding of the 238U resonances, increases
resonance absorption, and increases plutonium production per unit mass of
238U in the fuel. Therefore, even though uranium-thorium PWR fuel contains
5.5 times less 238U than the slightly enriched uranium fuel of Figure 3.1,
the reduction in plutonium generation is not nearly so great.
The discharge fuel from the national reactor could be shipped to an
internationally controlled centralized reprocessing center. Although the
plutonium could be allowed to follow the fission products to the high
level wastes, the reprocessing chemistry is such that this would not
materially simplify the separation operations. If the plutonium were to
follow the high-level wastes, the fissile content of those wastes would
be as high as 3 to 4 weight percent, which is much greater than that in
discharge fuel from uranium-fueled water reactors. Thus, the high-level
wastes would have to be safeguarded. Alternatively, the plutonium could
be recovered and consumed in on-site plutonium-burner reactors, as shown
in the overall flowsheet of Figure 7.2. Safeguards issues remain whether
plutonium is allowed to follow the wastes or is consumed in a reactor at
the international center. The effect of denatured-uranium thorium fueling
is to reduce the necessary power of the international plutonium-burner
reactor by a factor of 2.9 below that required with denatured uranium
fueling.
As compared with the low-enrichment uranium cycle of Figure 7.2,
the denatured uranium-thorium cycle has the advantage that a single
international reprocessing center could service a larger number of
national reactors, with only a relatively small total power of plutomum-
burner reactors at the international center. However the required uranium
enrichment capacity would be greater than in the case of the low-enrichment
uranium cycle of Figure 7.1. The enriched product, containing about 58%
7-3
-------
-International Energy Center
412 Mw
1
U - Pu Fueled
PWR
E= 30.4 Mw dayAg
Fuel Life - 3 yr
0 = 0.342
L =0.8
Pigford-Yong. 1977
Fission Depleted U
Products '0.2 Mg
U 10.4 Mg 1-32Mg 0.47%235U
0.47%235u
Pu 0.76 Mg
52.7% Fissile
t
UI0.7Mg
Pu0.97Mg
576% Fissile
Natural U
10.8 Mg
-National Reactor
U 270 Mg
0.83%235U
Pu 0.244 Mg
71.8% Fissile
U 26.6 Mg
0.83%235U
Natural U
137 Mg
U02
Fuel
Fabrication
U 28.1 Mg
3%235U
U 28.5 Mg
3%235U
Conversion
and Isotope
Separation
KDOOMw
U Fueled
PWR
E = 30.4 Mw dayAg
Fuel Life = 3 yr
0 =0.342
L=0.8
.Separative
Work
105 Mg
Depleted U
' 134 Mg
0.25% 235U
Fig. 7.1 Annual quantities for LWR cycle for international
safeguards, national reactors fueled with low enrichment
(denatured) uranium. (E=fuel exposure, n=overall thermal
efficiency, Incapacity factor)
-------
141 Mw
U-Pu Fueled
PWR
E = 30.4 Mw day/kg
Fuel Life =3yr
0 = 0.34
L= 0.8
Pigford - Yang, 1977
— International Energy Center
Fission Depleted
Products Uranium
U3.56Mg O.I3Mg 3.49Mg
OIK I
0.47%235U f
PuQ260Mg -J
52.9% Fissile
Pur ex
Repro-
cessing
Pu0.257Mg
U 3.66 Mg
Pu0.332Mg
58.4% Fissile
Fission
Products Thorium
0.90 Mg 18.5 Mg
Thorex-Purex
Reprocessing
Pu 0.079 Mq
£75% Fissile
U02-Pu02
Fuel
Fabrication
Natural U
3.70 Mg
U 5.74 Mg
68%233U
40%235U
PU 0.080 Mg
75% Fissile
Thl8.7Mg
I
U 5.68 Mg j
U02-Th02
Fuel
Fabrication
U 0.69 Mg
57.7% 235U
Natural U
85.3 Mg
Th 19.3 Mg
U6.30Mg
6.I%233U
9.8% 235U
Thl9.5Mg
Conversion
and Isotope
Separation
Separative
Work
89 Mg
Depleted U
184.2 Mg
0.25%235U
National Reactor
1000 Mw
U-Th Fueled
PWR
E = 33.4 Mw d
Fuel Life = 3yr
H = 0.34
L = 0.8
Fig. 7.2 Annual quantities for LWR cycle for international safe-
guards, national reactors fueledwith thorium and denatured uranium
(E=fuel exposure, n=overall thermal efficiency, Incapacity factor)
-------
235U in uranium, would have to be safeguarded until it is diluted with the
recycled uranium in the fuel fabrication facility. Also, the denatured
U-Th cycle requires far more complicated reprocessing and fuel refabrica-
tion operations. The technology base for this fuel cycle would require
further development and engineering scale-up before industrial-scale
operations could begin.
The flowsheets for equilibrium fuel cycles indicate less annual
make-up uranium ore for uranium-thorium fueling. However, additional
uranium ore is required to establish the equilibrium inventories in this
fuel cycle. When evaluated on the basis of 30-yr lifetime uranium ore
requirements for the same total system power, the uranium-thorium cycle
of Figure 7.2 requires about 38% less uranium ore than the uranium cycle
of Figure 7.1.
The relative power of the international piutoniurn-consuming
reactor can be reduced further by using thorium instead of natural
uranium as the fertile material for this reactor. The thorium is to
be blended with plutonium from national reactors, as shown in Figure 7.3.
Fissile uranium from thorium discharged from the international plutonium-
burning reactor becomes an additional source of fissile make-up for the
national denatured-uranium thorium reactors. Material quantities for the
plutonium-thorium international reactor of Figure 7.3 were calculated from
the data of Matzie (M2). This combination reduces the necessary power of
the international plutonium-burning reactor by a factor of 4.9 below that
of the simple uranium-piutoniurn cycles of Figure 7.1. The data shown
here for the equilibrium fuel cycles indicate a further savings in the
rate of consumption of uranium ore per unit of total electrical energy
produced, as compared with Figure 7.2. However, the ore savings are not
appreciable when accumulated over the reactor lifetime.
7.4 Denatured-Uranium-Thorium Cycle with National Pk'R_ and
International LMFBR
Fast breeder reactors under international control could also be used
as plutonium burners and as the source for the fissile uranium make-up for
national denatured uranium-thorium reactors. Portions of the breeder
blanket, such as part of the radial blanket, could contain thorium instead
of depleted uranium. The thorium blanket would be reprocessed along with
recycled uranium-thorium fuel from the national reactors. The recovered
uranium would be diluted with natural or depleted uranium prior to off-site
shipment as denatured uranium. It is likely that this concept could be
technically possible by modifying the blanket loadings for even the first
generation LMFBR's, which are expected to be started on plutonium.
A flowsheet of such a fuel cycle involving international breeder
reactors is shown in Figure 7.4. This has been calculated from the
characteristics of a commerical-scale LMFBR designed for possible intro-
duction in this century (P2). It has been assumed that all of the
breeding-gain fissile production of the breeder is drawn off as 233U, to
be used as fissile make-up for the denatured U-Th national reactors. As
a result, no fissile breeding gain is available from this breeder to start
up additional breeders, i.e., the effective doubling time for breeder
fissile inventory becomes infinite. In principle additional breeder
capacity could be introduced as needed, even when existing breeders
operate at zero breeding gain, by starting the new breeders with isotopi-
7-6
-------
•International Energy Center
I
84 Mw
Pu-Th Fueled
PWR
E = 33.4 Mw day/kg
Fuel Life = 3yr
H = 0.34
L= 0.8
TU i nr\ i» Fission
Th 1.90 Mg Pr0ducts Th
U 0.023 Mg Q.08 Mg 1.88 Mg
94% Fissile i 1
Pu 0.14 Mg _[ L
44% Fissile
Fission
Products Th
0.9 Mg 18.6 Mg
t t
Th 18.8 Mg
U5.72Mg
7.0%233U
3.8%235U
National Reactor
1000 Mw
Thorex-Purex
Reprocessing
PuO.I4Mg
Th 1.94 Mg
PuO.21 Mg
57 % Fissile
r
Pu02-Th02
Fuel
Fabrication
;x
'9
fc
j
Pu QOTSMg
75% Fissile
Thorex-Purex
Reprocessing
>2
n
U 0.023 Mg
94% Fissile
1
U
U02-Th02
Fuel
Fabrication
75% Fissile
1
1
5.66 Mg j
1
1
Th !9.34Mg
U 6.24 Ma
U-Th Fueled
PWR
E = 33.4 Mw day/kg
Fuel Life = 3yr
q = 0.34
L = 0.8
Pigford - Maeda, 1977
U0.62Mg
Thl.96Mg 56%235U
Natural U-
74.4 Mg
6.7%233U
8.9%
Thl9.5Mg
Separative
Work
77 5 Mg
1
I
235
U
j Depleted U
73 4Mg
0.25%23\l
Fig. 7.3 Annual quantities for LWR cycle for international safeguards,
national reactors fueled with thorium and denatured uranium, inter-
national reactors fueled with thorium and plutonium (E=fuel exposure,
n=overall thermal efficiency, Incapacity factor)
-------
CO
International
1674 Mw
i
LMFBR
U-Pu core
U + Th axial blanket
Th radial blanket
E = 68 Mw day/kg
0 = 0.42
L=0.8
Th Blanket
Fabrication
12.1 Mg
0.412 Mq
Energy Center -
Fission
Products
Fission
Products
Thorium
238U 149 Mg
l.99Mg
238U 16.2 Mg
Puf 2.03 Mg
• Thorium
12.7 Mq
0.91 Mg 290Mg
I 1
Thorex- Purex
Reprocessing
Puf
Q078Mg
U7.64
Mg
Fuel
Fabrication
Depleted
Uranium
!57Mg
U 7.91 Mg
II.4%233U
I 0 % 235U
Depleted Thorium j Th 17.7 Mg
Uranium |7.8Mg|
0.34 Mg i
-National Reactor-
U 7.31 Mg
6.9%233U
1.1% 235U
1000 Mw
Puf 0.079 Mg
Th !7.2Mg
U-Th Fueled
PWR
E=33.4 MwdayAg
Fuel Life = 3 yr
n = o. 34
L = 0.8
Pigford-Moedo-1977
Fig. 7.4 Annual quantities for national PWR fueled with thorium
9T?
and denatured uranium, international LMFBR produces U (E=fuel
exposure, n=overall thermal efficiency, Incapacity factor)
-------
cally enriched uranium. However, 23-sU start-up is not an economical
alternative for commercialized breeders (HI, P6). Therefore, operating
an international breeder as shown in Figure 6.4 would be possible only
after many decades when, even with an assumed zero growth of total
fission electric power, the assumed breeders have finally been intro-
duced to a level sufficient to replace the water reactors then being
retired. Earlier operation would entail a much higher relative power
from the international breeder, so that the breeder can then produce
additional fissile material for start-up of new breeders.
During the first few decades of breeder introduction the excess fissile
production from these breeders is likely to be needed to start-up new
breeders, so less fissile production is available as make-up for 23;jU for
national reactors. Consequently, the relative power of the international
breeder shown in Figure 7.4 is the minimum breeder power, relative to the
power of the national reactors, to supply the fissile make-up for the
national reactors. To maintain a finite breeder doubling time an even
larger relative power of the breeder would be required. However, the
necessary breeder power can be reduced somewhat if the more favorable
breeding gains calculated for future advanced breeders are assumed.
The fuel cycle flowsheet of Figure 7.4 has been calculated on the
assumption that thorium can be used in both radial and axial blankets.
The radial blanket alone will not produce sufficient 233U at this power
level. However, thorium in the axial blanket requires that the axial-
blanket thorium pellets be segregated from the core pellets prior to
reprocessing. Otherwise normal reprocessing of the entire fuel rod would
dilute the 233U with core uranium to the extent that it would be unsuitable
for use in the denatured-uranium cycle. However, if the breeder power were
increased to 3900 Mw, sufficient 233U would be produced in the separate
radial blanket, and normal head-end reprocessing techniques for core fuel
could be used.
Fuel cycles involving denatured thermal uranium reactors and
international breeders can provide excellent long-term ore utilization,
but they require the greatest total power and the greatest reprocessing-
refabrication capacity at the international facility. Also, such cycles
have all the complexity of reprocessing and refabrication facilities neces-
sary for both uranium-piutonium fueling and uranium-thorium fueling. They
appear to be the least realistic in terms of time schedule and availability.
7.5 National and International Fast Breeders
It is also technically possible for the national reactor to be a
breeder with a denatured 233U-uranium core, with the fuel discharged
from core and blanket sent to the international center for reprocessing.
The breeder at the international center would consume the plutonium
produced in the national breeder, and 233U produced in the international
breeder would be denatured by 238U dilution and exchanged for the plutonium
produced in the national breeder. However, full denaturing to ML ^y
in uranium may not be possible because of the high fissile concentration
in the breeder core required for criticality. Although the breeding gain
possible with 233U fueling in the breeder core is less than for plutomum
7-9
-------
Nationol Reactor-
•^ in ici i IUIIUMUI ciier
Fission
75 Mw
U-Pu Fueled
PWR
E =30.4 Mw day/kg
Fuel Life = 3 yr
q = 0.34
L = 0.8
Products
O.C
U 1.90 Mg
Puf 0.073 Mg
)7Mg
t
yy ueiiiei
Depleted
Uranium
l.86Mg
Purex
Repro-
cessing
Puf 0.072 Mgj
U l.95Mg
r"\ f\ i/""\^k A
PufO.I02Mg
i
uo2-
Fission
Products
0.90 Mg
_ , ^ j ^ , ,v
1
1
Thorium |
51.2
Thorex-Purex
Reprocessinc
Puf 0.0317 Mg
,
Pu02
Fuel
r— t , •
Fabrication
i
Natural U
l.97Mg
L
U02-Th02
Fuel
Fabrication
U 0.35 Mg
47|o/o235
Natural
Pigford - Moeda -Yang , 877
35.
U— »•
1 Mg
U
Conversion
and Isotope
Separation
1
'
1
*^\^
M(3 U7l8Mg
!0.3%2gU
I.48%235U
( ^ Puf 0.032 Mg
3 Th5l.7Mg
J 7 1 1 Mg |
1
I
Th52.5Mg
U7.39Mg
o "Jtn
9.8%253U
3.6%235U
1
"h 53.0 Mg |
1
1
i
Separative |
Work
36.4Mg
1000 Mw
U-Th Fueled
CANDU HWR
E= 16.0 Mw day/kg
ruel Life = 1.9 yr
q =0.30
L =0.8
Depleted U
^34.7
Mg
0.25%235U
Fig. 7.5 Annual quantities for national CANDU reactor fueled with
thorium and denatured uranium, international Pu-burning PWR (E=
fuel exposure, Coverall thermal efficiency, L=capacity factor)
-------
in the core, there are possible safety benefits due to smaller changes
in reactivity accompanying sodium voiding in the core. The international -
national breeder system does provide less ore consumption than a system
with thermal reactors at national sites, and thus it remains a possibility
for the very long-term.
7.6 Denatured-Uranium-Thorium Stowaway Cycle for HTGR
To avoid the safeguards issues of normal thorium fueling with 93%
235U make-up, as discussed in Section' 7.1, the HTGR design can 4>e
adapted to denatured-uranium thorium fueling. The flowsheet for the
near-equilibrium fuel cycle, without fuel reprocessing, has already
been shown in Figure 5.3. Resource requirements are listed in Table
5.4.
7.7 Denatured Uranium-Thorium Cycles with National Heavy-Water
Reactors
The pressure- tube heavy-water reactor, now commercialized in Canada
as the CANDU reactor, is another possibility for a national reactor to be
fueled with denatured uranium with or without thorium. The present cycle
of the CANDU reactor fueled with natural uranium, with storage of dis-
charge fuel, is the first possibility. Alternatively, this discharge fuel
could be reprocessed at an international facility and the recovered
Plutonium could be consumed in an on-site pi utoni urn-burner reactor.
However, because of the expense of reprocessing fuel with the low concen-
tration of plutonium formed in natural uranium fuel, the low burnup
reprocessing cycle will not become economical until natural uranium
prices become considerably higher than present contract prices. Instead,
the national CANDU could operate with slightly enriched uranium for more
economical reprocessing and more efficient resource utilization, quali-
tatively similar to the national reactor of Figure 7.1.
A CANDU national reactor could also be fueled with denatured uranium
and thorium, as shown in Figure 7.5. The reactor lattice is assumed to be
the same as in the present CANDU design. The fue exposure of 16 Mwday/kg
is that adopted in the study by Till and Chang (Tl), and their data have
been used to normalize the calculations for Figure 7.5. Because this fuel
burnup is more than twice as great as that in the natural uranium CANDU,
9% void volume has been provided for fission gasses (Tl). It is assumed,
for the purpose of these calculations, that no other J^l modifi cations
will be required. Even higher burnups would be expected for an optimum
fuel cycle involving fuel with fissile concentrations gr eater than natural
uranium. The national CANDU reactor of Figure 7 .5 must be operated for
many years with additional quantities of enriched 23fU before the
equilibrium conditions shown in this figure are attained.
As compared with the cycle of Figure 7. 2 involving the P»«sur12ed-
water national reactor, the CANDU reactor of Fl?"« J. 5 produces halfas
It reu s half th e
much fissile plutonium in the discharge fuel. It requ J^s a
of an international plutonium burner to consume the pluton urn,
as many CANDU national reactors can be served by a sl"9le ™^n^l°™Li-
plutonim-burner reactor. The lower plutonium generation f^he denatured
uranium-thorium CANDU reactor is a consequence of the more moderated
neutron spectrum and greater heterogeneity of the CANDU lattice.
7-11
-------
Because of the low neutron absorption in deuterium and the lack of a
pressure-vessel constraint in the CANDU reactor, sufficient heavy water
is used as moderator so that the ratio of epithermal flux to thermal
flux is much smaller than in the light-water reactor. This, together
with the greater spatial self-shielding of resonance neutrons in the
heavy-water lattice, results in less absorption in the 238U resonances.,
Because of the denaturing criterion of Equation (1). there are no large
differences in the ratio of 238U to fissile uranium in the PWR and CANDU
reactors of Figures 7.2 and 7.5. In either reactor the rate of plutonium
generation is proportional to the rate of absorption in 238U divided by
the fission rate. Therefore, because of the lower absorption in 238U
resonances in the CANDU, the plutonium generation in the U-Th CANDU is
over two-fold less than in the U-Th PWR. The lower burnup in the CANDU
does result in less plutonium consumption during irradiation than in
the PWR, but the lower plutonium production rate is dominant and results
in two-fold less plutonium in the U-Th CANDU discharge fuel.
The U-Th CANDU reactor of Figure 7.5 operates with an overall
conversion ratio of 0.90, as compared with 0.67 for the U-Th PWR.
Consequently more 233U is bred in the CANDU, and much less fissile
make-up and uranium ore are required. In both cases the make-up uranium
is of sufficient enrichment that the enrichment supply and fuel fabrica-
tion must be under the same international safeguards.
Because of the radioactivity of 1.91-yr 228Th and its daughters
in irradiated thorium, the recovered thorium must be stored for several
years before it can be recycled. Until such recycle occurs, the CANDU
requires 2.7 times more make-up thorium than the U-Th PWR, because of
the relatively low fuel exposure chosen for the CANDU and because of
the higher concentration of thorium in the CANDU fuel.
The flowsheet for a national U-Th CANDU reactor fueled with make-up
233U from an international breeder is shown in Figure 7.6. Even though
the U-Th CANDU produces over two-fold less plutonium for the breeder
fissile balance, it requires much less 23dU make-up because of its higher
conversion ratio. Thus the required power of the international breeder
for the U-Th CANDU cycle is 3.6 times smaller than that for the U-Th PWR
cycle of Figure 7.2. The same considerations as to the time scale of
feasibility of such a cycle, as discussed earlier for the U-Th PWR with
an international breeder, also apply here.
These calculations indicate that when compared with a pressurized-
water reactor, both operating on the denatured-uranium thorium fuel cycle,
the CANDU heavy water reactor:
(a) produces about two-fold less plutonium,
(b) requires about two-fold less natural uranium for the make-up
fuel and requires about two-fold less separative work if the
make-up fissile uranium is obtained by isotope separation,
(c) requires about three-fold greater amount of thorium make-up
fuel prior to thorium recycle, assuming the burnups used in
this analysis,
(d) requires about two-fold less power of an international plutonium-
burner reactor to consume the plutonium,
7-12
-------
co
International
465 Mw
i
Th3.75Mg
0.127 Mg
LMFBR
U-Pu core
U + Th axial blanket
Th radial blanket
E = 68 Mw dayAg
H = 0.42
L=0.8
Th Blanket
Fabrication
Energy Center -
Fission
Products
0.32Mg
238U 3.81 Mg
Puf 0.54 Mg
238U 4.15 Mg
Puf 0.57 Mg
•»- Thorium
3.95 Mg
Purex
Repro-
cessing
Fission .
Products Thorium'
0.91 Mg 54.6 Mg y 7.45 Mg
| 10.4 % 233U
I 0.66%235U
-Notional Reactor
Puf
1 I
Thorex- Purex
Reprocessing
Puf 0.0339 Mg
Fuel
Fabrication
Core and
U Blanket
0.0336 Mg
Th 51.4 Mg
I
Fuel
Fabrication
Depleted
Uranium
0.42Mg
r i
• U 7.62 Mg
II.6%233U
j0.64%235U
Depleted Thonumj Th52.2 Mg
Uranium 52.7 Mgj
0.19 Mg |
I
1000 Mw
U-Th Fueled
CANDU HWR
E = 16. OMw dayAg
Fuel Life = 1.9 yr
0 = 0.30
L = 0.8
Pigfor
-------
(e) requires about four-fold less power from an international breeder
if the fissile uranium make-up is obtained as excess 233U from
the breeder.
Our analysis is strictly limited to the considerations outlined
herein, and there is no intent to imply conclusions as to the superiority
of one reactor type over another. In further consideration of the CANDU
reactor as a candidate for a safeguarded national reactor, it would be
important to evaluate the required development of higher-burnup fuels
for the CANDU, the licensability of the CANDU under the same criteria
that are applied to light water reactors, the possible safeguards vul-
nerability of the CANDU because of its provision for frequent refueling
with small fuel batches, and the relative costs of the CANDU and PWR
systems.
Many of the features outlined above for the CANDU heavy water
reactor as a national reactor fueled with denatured uranium would also
apply to any other reactor of equivalent ratio, such as the more advanced
high-conversion-ratio modifications of the HTGR reactor listed in Table
5.3. Light water reactors designed to high conversion ratio are also
future possibilities (HI).
7.8 Enrichment Vulnerability of Denatured-Uranium Fuel
Using denatured uranium with a fissile content in the range of 10
to 20% creates a new safeguards issue in that relatively little work of
isotope separation would be required to isotpoically enrich this uranium
to the level of highly enriched material. This can be illustrated in
terms of the 235U equivalent. Highly enriched uranium is usually regarded
to be about 90% 235U, which is made by isotopically enriching natural
uranium. Of the total work required to enrich natural uranium to 90%
23t>U, about 90% of the work is expended in enriching to 20% 235U. Only
10% more work is required to further enrich to 90% 235U. This illustrates
the relative ease of making highly enriched uranium from uranium initially
containing as much as 10 to 20% fissile concentration.
Because of the lower atomic mass of 233U, the relative work required
to enrich 233y_238u denatured uranium to the high-enrichment level would
be even less than estimated above. Although recycled uranium containing
233U and z32U could not be enriched in commercial isotope separation plants
because of the radioactivity, there are many relatively small and not
necessarily efficient isotope separation systems that could enrich this
uranium. The technology to carry out such enrichment on non-economical,
non-commercial scale is available in the open literature. This is another
aspect of the denatured uranium-thorium cycle that requires further
evaluation.
To illustrate, the relative amounts of separative energy and plant
capacity to produce 90% fissile uranium are estimated for the following
two reactor fuels:
(a) normal PWR fuel, containing 3% 235U in 235y_238y
(b) denatured uranium fuel containing 12% 233U in 233U-238U.
For purposes of this illustration, we assume ideal close-separation
cascades and 0.3% enrichment tails. The total interstage flow J^per
7-14
-------
E1ttf^«Mfrfl1b/(M)?d t0 ^ 1dea' Separat1°" '"tor . and
J =
(7-2)
where
= 2x-l £n ~
1-x (7.4)
x = atom fraction of light isotope
and F, P, W denote feed, product, and tails, respectively. Also, for
the close separation of heavy isotopes:
(7-5)
where A. and A. are the atomic weights of the heavy and light species,
respectively. JFrom these equations and assumed compositions, we calculate
that
J for 235U.238U =
J for 233U.238U ' (7_8)
Assuming that energy requirement and necessary equipment capacity are both
proportional to the total interstage flow for a given method of separation
(B6), about eight-fold less energy and plant capacity are required to
enrich the 2d3y_238u fue] t0 gg% product than are required for the 3%
235y_238y fue] normally used in pressurized water reactors.
7.9 Comparison of Denatured-Uranium Fuel Cycles
The principal fuel cycle quantities for the various denatured fuel
cycles are compared in Table 7.1. The uranium resource requirements,
presented here in terms of the annual quantities of contained elemental
uranium, are calculated for the equilibrium fuel cycles and do not
reflect the additional non-equilibrium start-up requirements. If it is
assumed that the international piutoniurn-burner reactor operates at a
power level of 1000 Mwe, the last column represents the total electrical
generating capability, in Gwe, of the system of national reactors and
the international reactor which serves them. The annual resource
requirements per unit of total system generating capacity are smallest
when the international reactor is a breeder. Without breeders the smallest
annual uranium requirement occurs for CANDU national reactors. The net
thorium consumption by these reactors can be made small by later recycling
the stored thorium.
A national denatured uranium-thorium PWR would reduce plutonium
generation by only a factor of 2.9, and it could achieve a modest saving
in uranium resources. Greater reductions in plutonium generation and in
uranium-resource requirements are possible with a plutonium-thorium inter-
7-15
-------
Table 7.1 Comparison of Fuel Cycle Quantities for Denatured Fuel Cycles
National Reactor
Type/Fuel
PWR
Denatured U
(31;, 235u)
PWR
Denatured U
(3% 235U)
PWR
Denatured U
+ Th
PWR
Denatured U
+ Th
PWR
Denatured U
+ Th
PWR
Denatured U
+ Th
HTGR
Denatured U
+ Th
CANDU
Denatured U
+ Th
CANDU
Denatured U
+ Th
International Pu-
Burner Reactors
Type/Fuel
none
PWR
Pu + Natural U
PWR
Pu + Natural U
PWR
Pu + Th
LMFBR
Pu + depleted-U core
Th + U axial blankets
Th radial blanket
LMFBR
Pu + depleted-U core
Th + U axial blankets
Th radial blanket
none
PWR
Pu + Natural U
LMFBR
Pu + depleted-U core
Th + U axial blankets
Th radial blanket
Annual Resource Requi rements-
Mg/Gwe yr
U
168.7^
104.7^
78.0^
68.6^
0.714-/
0.815^
36^
34. ^
0.41 6^
Th£/
.
.
17.1
19.8
11.4
9.67
3.76
49.3
38.7
Separative
Work
108
74.4
78.0
71.5
-
-
132
33.9
-
Th^/
Storage
.
_
16.2
17.2
10.8
9.17
3.46
47.6
37.3
Relative Electric Power
National Reactors
International Reactors
2.43
7.09
11.9
0.60
0.26
13.3
2.15
Total Systems
International Reactors
3.43
8.09
12.9
1.60
1.26
14.3
3.15
— Calculated for equilibrium fuel cycle, total electrical generating capacity of system of international and national reactors = 1000 Mwe
- Natural uranium as U308 from milling and conversion of uranium ore
— Natural thorium (232Th), as ThO? from milling and conversion of thorium ore
- Stored thorium can be recycled after storage for about 4 to 17 years
- Depleted uranium stockpiled from isotope separation
-------
national reactor and with a national CANDU reactor fueled with denatured
uranium and thorium.
It may be possible to turn to advantage the fact that all the
denatured-uranium fuel cycles considered in this study require basically
the same type of institutional and political agreements. In all cases
the national reactors receive qualitatively similar denatured fresh fuel,
and all national reactors discharge fuel containing unused energy resources,
including enough plutonium to require that the discharged fuel be safe-
guarded. What changes from one cycle to another are the detailed facilities
at the international sites. Therefore, if appropriate institutional and
political agreements can be negotiated to make possible even the simplest
of the cycles, i.e., the international stowaway fuel cycle, then substan-
tially the same agreements and arrangements can remain in effect as more
and more resource-efficient fuel cycles are introduced in the course of
time. Thus it is important to fully analyze such safeguards fuel cycles for
their economic, social and political consequences as well as their technical
viability (U2).
7-17
-------
8> Radioactivity, Long-Term Toxicity. and Artist*
High-Level Radioactive Waste? ~
8.1 Introduction
Here we present a comparison of the radioactivity and long-term
toxicity of high-level radioactive wastes from the fuel cycles dis-
cussed in Chapters 3, 4, and 5. Emphasis is given to the actinides in
the high-level wastes, which control the toxicity of these wastes
after the fission product period of a few hundred years. For a more
detailed comparison the transuranic wastes from fuel reprocessing '
and refabrication should also be considered, because the amount of
actinide activity in these wastes is likely to be comparable to that
in the high-level wastes (HI, P7, P8).
The waste toxicity considered here is the ingestion toxicity,
defined as
ingestion toxicity = > Rrl (8.1)
where N. is the number of atoms of nuclide 1_ in the wastes at any time
Jt, A. is the radioactive decay constant, and RCG^ is the radioactivity
concentration guide for ingestion of nuclide _i_, for unrestricted expo-
sure to the public (U3).
It is to be emphasized that the ingestion toxicity of wastes,
here presented on the basis of quantities per gigawatt-year of
electrical energy generation, is only a crude and limited index of
possible hazards of radioactive wastes. It does not take into
account the long-term integrity of the waste form or the differences
in transport of the different waste elements through the emplacement
medium and through the environment.
8.2 Radioactive Wastes From the Reference U-Fueled Light-Water Reactor
The radioactivity of plutonium, americium, and curium in the high-
level reprocessing wastes for the uranium-fueled water reactor of
Figure 3.2 are shown in Figures 8.1 and 8.2 (P2). These quantities
are calculated for the amount of wastes generated by reprocessing the
fuel discharged yearly by a 1000 Mw reactor. The amount of 238Pu in
the high-level wastes increases with time because of the decay of
2lt2mAm and 2k2Cm, the amount of 21t0Pu increases because of the decay
of ^Cm and the amount 239Pu increases because of the decay f43Am
and 21t3Cm. The principal contributors to the long-term ingestion
toxicity of these wastes are shown in Figure 8.3 (Bl, HI, PI). During
the first 600 years of waste storage the ingestion toxicity is dominated
by 9°Sr in the fission products. Thereafter, 21tlAm and 2l+3Am are most
8-1
-------
CO
I
no
i 10 io io3 io4 ios io6
DECAY TIME. y«r«
0.1
0 I
-------
10 D2 D3 D4 I05 I06
Storage Time, years
I07 D8
Flj. 8.3 Ingestlon toxldty of high-level wastes
from U-fueled PUR (33 HN day/kg, 0.5% U and Pu lost
to waste)
8-3
-------
Depleted Uranium
from Isotope
Separation
i i I i I
10 I02 I03 I04 I05 I06 I07 I08
Storage Time, years
Discharge U Fuel
from LWR
LWR With Self-generated
Pu Recycle, 0.5% U and~
in Wastes
U-Th
Fueled HTGR
U Fueled LWR,
Q5 % U and Pu in Wastes
10
I02 O3 I04 KD5
Storage Time, years
Fig. 8.4 Relative Ingestlon toxlcity of fuel cycle
residuals from U-fueled PHP. (33 Mw day/kg, 0.5X U and
Pu lost to waste, 51 lost to mill tailings, 0.25* 235U
In depleted uranium)
Fig. 8.5 Ingestlon toxlcltles of high-level wastes
from various fuel cycles (33 Nw day/kg for LMR's.
95 Mw day/kg for HTGR)
8-4
-------
important, followed by 239Pu and 2it0pu and then b 226Ra 225
Si. 26?a aPPears from the decay of 23"U, 238Pu, 2^m, and
2l+2Cm initially in the high-level wastes, and 225Ra 1s formed
from the decay of 2"ipu, ^Am, and 23>! After ^ ?06 years
of storage, the waste toxicity decays to a level due to 226Ra in
secular equilibrium with the small amount of 238U in the high-
level wastes. Although the long-lived fission product 129I will
be recovered separately from the bulk of the fission products con-
taining the actinides, its long life and high toxicity require its
inclusion in an overall toxicity analysis.
8.3 Waste Toxicities in Perspective
The ingestion toxicities for the high-level wastes from repro-
cessing non-recycled uranium fuel are compared with the toxicities
of other residuals from this same fuel cycle in Figure 8.4 (Bl,
H4, P7). These toxicities are normalized to that of the uranium
ore mined for one gigawatt year of reactor operation. The ore
toxicity is due mainly to 226Ra, which is in secular equilibrium in
the 238U decay chain. In the processes of milling and concentrating
uranium ore 226Ra and its precursor 80,000-yr 230Th follow the
tailings. Therefore, the ore toxicity is preserved in the mill
tailings for a few hundred thousand years until 230Th decays. There-
after the tailing toxicity continues at a lower level determined
by the residual uranium in the tailings, assumed here to be 5% of
the uranium processes. If the depleted uranium from isotope separa-
tion is never used for breeder fuel, the uranium daughters, par-
ticularly 226Ra, in this stored UFg will eventually be restored to
a toxicity level with a few percent of the original ore toxicity.
The toxicity of the high-level wastes falls below that of the
original ore after a period~of about 600 years. The total toxicity
of all residuals falls below that of the original uranium ore after
a decay time of about 140,000 years. This minimum results from
the enrichment of natural 23I+U in isotope separation and its destruc-
tion in the reactor by neutron absorption, thereby depleting one of
the sources of 226Ra.
The toxicity indices are not measures of hazards, in part because
they take no account of the barriers which isolate these wastes from
the biosphere nor the behavior of different radioactive elements with
respect to these barriers. However, the longer-term toxi cities of
the high-level reprocessing wastes are due to radium, which is the
same element that controls the ore toxicity. The long-term radium
toxicity of the reprocessing wastes is considerably less than the
radium toxicity of the ore. It seems reasonable that radium ultimately
appearing in the high-level wastes can be geologically isolated so that
the waste material has less access to the environment than the radium in
the natural ore.
A comparison of the hazards from high-level wastes and uranium ores
can be derived from the results of Burkholder, £t al_ (B8, B9), who have
analyzed the long-term migration of fission products, actinides, and de-
cay daughters from a model geologic repository; with sorption retardation
of individual radionuclides according to chemical species. Hazards in
8-5
-------
terms of fifty-year integrated individual doses were calculated for
migration times from 102 to 107 years after emplacement. Hazards from
americium and plutonium were found to be less than the longer-term
hazard from radium in the high-level waste. For migration pathways
through the geologic medium as great as 480 meters the hazard from 226Ra
was found to be greater than the hazard from 90Sr, the fission product
which dominates the fission-product toxicity curve (Fig. 8.3) during the
first few hundred years. The only fission products found to present
greater hazards than 226Ra were 9§Tc and 129I, and then only for the long-
est geologic pathways and for relatively rapid leaching (0.3%/yr) of tech-
netium and iodine from the wastes. Therefore, it is important to recog-
nize that:
(a) the principal hazard from migration of radionuclides from high-
level waste in geologic isolation may result from the long-term
migration of 22gRa, the same radionuclide that controls the in-
gestion and migration hazard from the original uranium ore and
from the uranium mill tailings, and
(b) the amount of 226Ra in the high-level wastes from reprocessing
uranium fuel is less than the amount of 226Ra in the ore mined
to create these wastes (cf. Fig. 8.4).
Burkholder's (B8) analysis of hazards from radionuclide migration (B8),
which assumes a ground water .velocity as high as 110 meters/year, pro-
vides data on the effect of migration distance upon the 50-year dose
from 226Ra. Increasing the necessary migration distance from 160 meters,
as might be representative of a shallow ore body, to as much as 16,000
meters, as might be obtainable in a geologic isolation, decreases the
50-year 226Ra dose by a factor of twenty or more. Much larger attenua-
tions occur for most other radionuclides. These 226Ra doses are relative-
ly insensitive to the dissolution rate of the radioactive source material,
over a wide range of dissolution rates from 0.003 to 0'.3%/yr.
These data for a model repository illustrate that high-level waste
emplaced in a geologic repository, with sorption and transport properties
representative of this model repository, may be expected to result in less
actual hazard from nuclide migration than the hazards which would other-
wise result from the ore body which produced these wastes; assuming that
both of these sources of 226Ra are exposed to the same mode of groundwater
transport.
8.4 Effect of Pu Recycle on High-Level Haste Toxicity
Toxicities of high-level wastes from a light-water reactor with and
without plutonium recycle are compared in Figure 8.5. Recycling plutonium
increases the production of americium and curium (P2), whose radioactivity
and decay daughters increase the ingestion toxicity byaboutan order of
magnitude during the period governed by actinides and 226Ra.
8.5 Toxicity of Unreprocessed Uranium Fuel
As shown in Figure 8.5, the actinide toxicity of unreprocessed ura-
nium fuel from a light water reactor, which contains all of the plutonium
discharged from the reactor, is about fifty times greater than the toxicity
of wastes from uranium fuel which has been reprocessed for recovery of
uranium and plutonium. This conclusion applies to the period from one
thousand to one million years.
8-6
-------
00
I 10 KT10* I04 1C
Storoge Time, years
10° 10'
10 I02 I03 K>4 I05
Storage Time, years
10° 10'
Fig. 8.6 Pu radioactivity In high-level wastes from
Z35U-Th-fueled PWR with U recycle (33.4 HN day/kg.
l.St Th and U lost to waste)
Fig. 8.7 Actlnlde radioactivity In high-level wastes
from Z35U-Th-fueled PUR with U recycle (33.4 HN day/kg,
l.St Th and U lost to waste)
-------
8.6 High Level Hastes from the PWR Fueled with 235U. Th. and Recycled U
The radioactivity of plutonium radionuclides in the high-level re-
processing wastes for the equilibrium fuel cycle of the 235U-Th fueled
PWR are shown in Figure 8.6, calculated on the basis of all plutonium
in the discharge fuel following the high-level wastes. The elemental
radioactivity of the actinides and their daughters are shown in Figure
8.7. As compared with the plutonium in high-level wastes from reproces-
sing uranium fuel, the 235U-Th PWR wastes contain over 100 times more
238Pu, about the same quantities of 23*Pu, 240Pu, and 237Np, and 103 to
101* times 101* times less Am and Cm.
The ingestion toxicity of these thorium-cycle wastes is shown in
Figure 8.8. Comparing with Figure 8.3 for the uranium-fuel wastes, the
smaller amounts of Am and Cm in the thorium-cycle wastes result in rela-
tively low waste toxicity after the fission-product period of about 600
years. The uranium activity and toxicity in these thorium-cycle wastes
is relatively large because of the 232U, 233U, and 234U in the recycled
uranium, a fraction of which is lost to the wastes in each reprocessing
cycle. The 234U and 238Pu result ultimately in the relatively large tox-
icity peak for 226Ra.
The toxicity of the waste residuals from uranium and thorium milling
for the U-Th-fueled PWR, as well as high-level reprocessing wastes, are
plotted versus storage time in Figure 8.9. The presence of 230Th in the
natural thorium greatly increases the long-term toxicity of the milling
residuals. At a concentration of 100 ppm of 230Th, the 226Ra daughter of
230Th dominates the toxicity of the thorium ore. The 226Ra remains with
the tails from thorium milling, but it disappears by decay after about
10,000 years. Thereafter the toxicity of the thorium mill tailings reaches
the level due to 230Th-226Ra in the residual thorium. A loss of 5% of
the thorium to the tailings has been assumed.
With 100 ppm 230Th in thorium ore, the toxicity of the thorium tail-
ings is greater than that of the uranium tailings, both on the basis of
fuel cycle quantities per unit of energy produced as well as on the basis
of equal quantities of heavy element recovered. If the thorium is free
of 230Th,the thorium tailings have a lower ingestion toxicity than do the
uranium tailings. The early toxicity of the tailings from pure 232Th is
due to 5.75-yr 228Ra and its daughters. This decays after a few decades
to the toxicity of the 5% residual 232Th and its daughters in the tailings.
8.7 High Level Wastes from the PVJR Fueled with Pu, Th, and Recycled U
Wastes from thorium fueling with plutonium make-up and uranium re-
cycle include the high concentrations of 238Pu, 232U, 236U, and 237Np
and their daughters resulting from the recycle of bred uranium as well as
the high concentrations of americium and curium and their daughters re-
sulting from plutonium irradiation. With plutonium make-up there is in-
centive to recover and recycle the plutonium remaining in the dishcarge
fuel, and such recycle has been assumed in the fuel burnup calculations
for this cycle. Therefore, for the purpose of waste calculations it is
assumed that 1.5% of the plutonium and uranium in the discharge fuel is
lost to the high-level wastes. The activities of plutonium radionuclides
in the high level wastes from the equilibrium Pu-Th PWR fuel cycle are
shown in Figure 8.10. The elemental activities are shown in Figure 8.11
and the ingestion toxicities are shown in Figure 8.12.
8-8
-------
10
oo
i
10
10*
Storage Time, years
10° 10'
10
10
10" Da O* "i?"
Storage Timt, y«on
»'
Fig. 8.8 Ingest ion toxldty of high-level wastes
from 235U-Th-fueled PWR with U recycle (33.4 HN day/kg,
1.5t Th »nd U lost to waste)
Fig. 8.9 Ingestlon toxlcity of fuel cycle residuals
from 235U-Th-fueled PUR with U recycle (33.4 (fa day/kg,
l.SX Th and U lost to Haste. SS lost to Hill tailings)
-------
CO
—I
o
I K) I02 I03 I04 I05 Kf
Storage Time, years
10'
10 K>2 10* O4 IOS 10
Storage Time, years
,6
10'
Figure 8.10 Pu radioactivity In high-level wastes from
Pu-Th-fueled PUR with U and Pu recycle (33.4 NX day/kg,
1.51 Th, U, and Pu lost to waste)
Fig. 8.11 Actlnlde radioactivity In high-level nstes
from Pu-Th-fueled PM) with U and Pu recycle (33.4 M»
day/kg. 1.51 Th, U. and Pu lost to Haste)
10 I02 10* I04 IOS I06 I07
Storage Time, yean
Fig. 8.12 Ingestlon toxldty of high-level wastes fron
Pu-Th-fueled PVR with U and Pu recycle (33.4 H» day/kg.
1.5S Th, U. and Pu lost to waste)
-------
As a result of the increased americium, curium, and plutonium
the minimum in the waste toxicity at about 1000 years previously
noted for 235U-Th cycles does not occur. As shown in Figure 8 12,
the effect of the plutonium make-up is to raise the waste toxicity
by about sixty fold after the high-toxicity fission products have
decayed. During the period of the 225,226Ra peaks^ the waste
toxicity is essentially the same as for 235U make-up.
The unusually high radioactivity of Np until about 100,000
years, as shown in Figure 8.11, is due to 239Np in secular equilibrium
with 2H3Am. After Am decays the Np activity relaxes to the longer-
term level due to 237Np.
If all the plutonium in the discharge fuel were allowed to go
directly to the high-level wastes there would be an increase by a
factor of 67 in the initial activities of 239Pu, 21+0Pu, and 21+1Pu,
in Figures 8.10 and 8.11 and in the plutonium toxicity in Figure 8.12.
However, the total waste toxicity is affected appreciably by plu-
tonium only during a time interval at about 10,000 years of decay.
At this time the important plutonium radionuclides are 239Pu and 21*°Pu.
Most of the 239Pu and much of the 2tt°Pu will have appeared from the
decay of americium and curium in the wastes rather than from the plu-
tonium initially in the wastes. Allowing all the plutonium to
follow the wastes would not cause a significant increase in the
total toxicity of these high-level wastes. Therefore, the main
effect on waste toxicity resulting from choosing plutonium as fissile
make-up in the Pu-Th-U cycle is the increased production of americium
and curium.
8.8 High-level Wastes from the Uranjurn-Fueled and Thorium-Fueled
Heavy-Hater CANDU Reactors
The actinide radioactivity and the ingestion toxicity of the
discharge fuel from the uranium-fueled CANDU reactor are shown in
Figures 8.13, 8.14, and 8.15. The activities of 239Pu, 2l+0Pu, and
21tlPu are over 200-fold greater than in the high-level wastes from
the uranium-fueled PWR (cf. Figure 8.1), because the CANDU fuel has
not been reprocessed for plutonium recovery. The initial activity
of americium in the CANDU fuel is about the same as in the U-PWR high-
level wastes, but it increases about 10-fold in the first 100 years
of storage, due to the decay of 241Pu in the CANDU fuel. The total
ingestion toxicity of the discharged CANDU fuel is comparable to that
of unreprocessed PWR fuel shown in Figure 8.5.
The radioactivity and ingestion toxicities of high-level re-
processing wastes from the equilibrium fuel cycle of the 235U-Th-
fueled CANDU reactor, operating with uranium recycle, are shown in
Figures 8.16, 8.17, and 8.18. The quantities are quite similar to
8-11
-------
CO
ro
I 10 IO2 10" 10'' to"
Storage Time, years
iO6 .O7
Fig. 8.13 Pu radioactivity in natural-u-fueled CANOU
reactor discharge fuel (7.5 MM day/kg)
I 10 I02 IO3 IO4 IO5 IO6 IO7
Storage Time, years
F1g. 8.14 Actlnide radioactivity in natural-U-fueled
CAHOU reactor discharge fuel (7.5 MM day/kg)
I
10
I 10 I02 10* I04 I05 10* IOT
Storage Time, years
F1g. 8.15 Ingestlm taxlclty of natural-U-fuel«d CMDU
mctor discharge fuel (7,5 N* day/kg)
-------
00
I
co
10 10'
10'
Storage Time, years
Fig, 8.IS Py r«
-------
those in Figures 8.6, 8.7, and 8.8 for the 235U-Th-fueled PWR.
The peak amount of 225Ra in the stored wastes is greater for the
CANDU fuel cycle because of the greater throughput of 233U in the
reprocessing cycle, resulting in larger quantities of 233U lost to
the wastes.
The radioactivity and ingestion toxicities of high-level re-
processing wastes from the equilibrium fuel cycle of the Pu-Th-
fueled CANDU, operating with U-Pu recycle, are shown in Figures
8.19, 8.20, and 8.21. All actinide quantities, except for 233U,
are smaller than in the case of the Pu-Th-fueled PWR.
8.9 High-level Wastes from the Reference 235U-Th-Fueled HTGR
The actinide radioactivity and ingestion toxicities of high-
level reprocessing wastes from the equilibrium fuel cycle of the
reference 235U-Th-fueled HTGR, operating with uranium recycle and
without reprocessing cross-over, are shown in Figures 8.22, 8.23,
and 8.24. As compared with the waste properties for the 235U-Th-
fueled PWR, shown in Figures 8.6, 8.7, and 8.8, the HTGR wastes
contain much greater radioactivity quantities of U, Np, Pu, Am,
and Cm than do the PWR wastes. This is a consequence of the much
higher burnup of the HTGR fuel cycle. Actinide cross-over in HTGR
fuel reprocessing results in a small increase in the activities
and toxicities of the HTGR wastes.
The total ingestion toxicity of the high-level wastes from
the 235U-Th HTGR is compared with that of other fuel cycles in
Figure 8.5. The curve of the HTGR wastes is typical of that for
any of the 235U-Th fuel cycles. As has been explained-in Section
8.6., the relatively small amounts of Am and Cm in 235U-Th fuel
cycle result in relatively little waste to xi city during the period
of 103 to 105 years, after the fission products have decayed.
However, the relatively large 226Ra peak at 2xl05 years for the
HTGR 235U-Th fuel cycle brings the toxicity of these wastes to the
level of unreprocessed uranium fuel from PWR's.
Ingestion toxicities of long-term residuals from the HTGR
235U-Th fuel cycle are shown in Figure 8.25. As compared with the
similar plot (Figure 8.9) for the PWR 235U-Th fuel cycle, the lower
ingestion toxicity of the HTGR thorium mill tailings reflects the
lower consumption of thorium in this fuel cycle, a consequence of
the higher irradiation exposure of HTGR fuel. These differences in
thorium consumption and thorium mill tailing toxicities become much
smaller if thorium is recycled. The time trends of the toxicity
of thorium mill tailings are explained in Section 8.6.
8-14
-------
00
I
_«J
en
10 10
Storage Time, years
Fig. 8.19 Pu radioactivity 1n high-level wastes from
Pu-Th-fueled CANDU reactor with U recycle (27 Mu day/kg,
0.51 Th and U lost to waste)
Storage Time, years
Fig. 6.20 Actlnlde radioactivity In high-level wastes
from Pu-Th-fueled CANDU reactor with U recycle (27 Hw
day/kg, 0.51 Th and U lost to waste)
Storage Time, years
Fig. 8.21 Ingestlon toxldty of high-level wastes from
Pu-Th-fueled CANDU reactor with U recycle (27 Hw day/kg,
0.51 Th and U lost to waste)
-------
CO
10"
Storage Time, years
10=
10'
Fig. 8.22 Pu radioactivity In high-level wastes fro»
Z35U-Th-fueled HTSR with U recycle (95 MM day/kg.
0.7SS U and Th lost to waste)
KT 10" 10"
Storage Time, years
10"
Fig. 8.23 Actlnlde radioactivity In Mgb-level nastts
from Z3SU-Th-fueled HTSR with U recycle (95 At day/kg.
0.7SX U and Th lost to waste)
-------
00
I 10 I02 I03 IO4 I05 I06 I07
Slorogt Time, years
Th Mill Tailings
(with noMOTh)
10
10
Storage Time, years
F1g. 8.24 Ingestton toxldty of high-level wastes fron
235U-Th-fueled HTSR with U recycle (95 Mw day/kg,
0.7SX U and Th lost to waste)
Fig. 8.25 Ingestlon toxldty of fuel cycle residual!
from 235U-Th-fueled HTGR with U recycle (95 HM day/kg.
0.75X U and Th lost to waste)
-------
8.10 Comparison of Actim'de Sources in High-level Wastes
From Alternate Fuel Cycles
The previous plots have shown that the hazard potential,
i.e. the toxicity, of the high-level wastes is dominated by
fission products for the first few hundred years, followed by
2ttlAm and 21+3Am, then by 239Pu and 240Pu, and finally by 226Ra
and 225Ra. The quantities of actinide precursors of each of these
radionuclides are summarized in Table 8.1 for several alternate fuel
cycles. For the purpose of comparison 0.5% of the recycled actinides
are assumed to be lost to the high-level wastes.
Adopting the high level wastes from reprocessing uranium fuel
from the pressurized-water reactor as a reference for comparison,
the relative quantities of actinide precursors in the wastes from
the other fuel cycles are characterized as follows:
1. The unreprocessed PWR discharge fuel will ultimately
contain about 20 times more 21+1Am, the same quantity of 2k3l\m,
about 50 times more 239Pu, 21f0Pu, and 226Ra, and about twice as
much 225Ra.
2. The high-level wastes from the U-Pu-fueled PWR with self-
generated Pu recycle will ultimately contain about 4 times more
24lhm, 9 times more 21|3Am, 7 times more 239Pu, 14 times more 2l+0Pu
5 times more 226Ra, and about the same quantity of 225Ra.
3. The high-level wastes from the 235U-Th-fueled PWR with self-
generated uranium recycle will ultimately contain about 10 times less
24ll\m, about 760 times less 243Am, about 6Q% more 239Pu, about 16%
less 2lt°Pu, 42 times more 226Ra, and about the same amount of 225Ra.
4. The high-level wastes from the Pu-Th-fueled PWR, with U-Pu
recycle, will ultimately contain about 23 times more 21tlAm, 37 times
more 2l+3Am, about 25 times more 239Pu, 21+0Pu, and 226Ra, and about
80% more Ra.
5. The high-level wastes from the 235U-Th-fueled HT6R with
self-generated uranium recycle will ul-timately contain about 3 times
less 241Am, 16 times less 243Am, half as much 239Pu and 2t*°Pu, over
30 times more 226Ra, and about half as much 225Ra.
6. The high-level wastes from the U-Pu-fueled LMFBR will
ultimately contain about 3 times more 2itlAm, one-third less 2i+3Am,
3 times more 239Pu, 2 times more 2l+0Pu, equal quantities of 226Ra,
and about half as much 225Ra.
8-18
-------
It has been concluded elsewhere (HI) that, in view of the
anticipated efficacy of geologic isolation, the range of about 50
or less in the potential actinide hazards of the high-level wastes
for the various fuel cycles considered herein does not appear to
present a strong incentive for choosing one fuel cycle over another.
If it is assumed that the possible diversion and misuse of
concentrated fissile material is to remain a long-term safeguards
issue, as has been discussed in Chapter 7, then attention must be
given to the long-term vulnerability of the appreciable quantities
of fissile plutonium and fissile uranium in these radioactive wastes.
The national-international safeguards fuel cycles discussed in
Chapter 7 require the premise that, because of the intense radio-
activity of fission products in discharge fuel, the plutonium
in discharge fuel is sufficiently self-protected and suitable for
storage at and shipment from dispersed national sites. Similar
logic would apply to high-level wastes containing fissile actinides,
as well as unreprocessed discharge fuel, during the first few
hundred years of storage or disposal, while the fission products
remain. Thereafter, the radioactivity of the wastes, per unit mass
of plutonium contained in these wastes, is actually less than that
of plutonium separated at the time of reprocessing. The 238Pu and
2ltlPu, which are the main contributors to plutonium radioactivity at
the time of reprocessing, will have decayed away in these wastes
after a few hundred years. Therefore, the fissile content of these
wastes ultimately exists in a relatively non-radioactive environment.
Because of the radioactive decay of 243Am, the amount of 239Pu
in the high-level reprocessing wastes increases with time, as has
been demonstrated by the four-fold increase illustrated in Figure
8.1 for the reference high-level wastes and shown also in Table 8.1.
Also, the fissile isotopic concentration in the plutonium present
in these wastes, after the time period for americium decay, is
greater than for plutonium recovered from discharge fuel.
The concentration of elemental plutonium in the reference
uranium-fuel high-level reprocessing wastes, after americium decay,
will be about 0.1 weight percent, assuming a four-fold dilution of
the fission products and actinides by borosilicate glass. This
compares with 0.94% for plutonium in discharged uranium PWR fuel.
The chemical technology which can recover plutonium from discharge
fuel in the presence of intense radioactivity can be reasonably
expected to recover plutonium from the relatively non radioactive
high-level waste mixture after americium decay. Whether geologic
isolation of these wastes for the purpose of environmental protection
of future generations, and whether such safeguards issues for future
generations are indeed relevant, are issues which may warrant further
consideration.
8-19
-------
Data in Table 8.1 show that the near-term inventory of fissile
Plutonium in unreprocessed discharge fuel is 200 times greater
than that in the high-level reprocessing wastes at the time of
reprocessing, and about 50 times greater when compared after storage
long enough for americium decay. If short-term or long-term safe-
guards of fissile inventory in stored discharge fuel or in stored
reprocessing wastes are important issues, then reprocessing to re-
cover and consume the plutonium by recycle may be indicated (M5).
However, the process of plutonium utilization increases the quan-
tities of fissile plutonium ultimately in the wastes, as is illus-
trated in Table 8.1 for self-generated plutonium recycle in light
water reactors and in fast breeders. The reduction ratio, i.e.,
the ratio of fissile plutonium inventory in discharge uranium fuel
from PWR's to that ultimately appearing in high-level reprocessing
wastes is about 10 for the recycle of plutonium in light-water
reactors and is about 23 for utilization of plutonium to start
first-generation fast breeders. The latter case is calculated on
the basis of the breeder start-up requirements shown in Table 6.1
and the breeder waste inventories shown in Table 8.1. The reduction
in plutonium inventory of wastes by recycling is much greater for
the near term before americium decay. Data in Table 8.1 indicate
a near-term reduction ratio of 140 for the PWR and 20 for the fast
breeder.
If the fissile inventories in stored discharge fuel and in
high-level reprocessing wastes are considered to be important short-
term or long-term safeguards issues, then the non-reprocess ing fuel
cycle would clearly be the least favorable.
8-20
-------
TABLE 3.1
Comparison of Actinide Quantities in High-level Wastes from Alternate Fuel Cycles (Basis = 1 Qw yr of reactor operation, L = 0.8,
150 days preprocessing cooling, quantities calculated at time of reprocessing).
CO
ro
U-fueled
PUR a.b/
Sources of •"•'Am, g atoms
Pu - 241
Am - 241
Cm - 245
Total
Sources "of 21l3Am, g atoms
Am - 243
Sources of 239Pu, g atoms
Pu - 239
Am - 243
Cm - 243
Total
Sources of 21*°Pu, g atoms
Pu - 240
Cm - 244
Total
Sources of 2?6Ra,i/ g atoms
U - 234
Pu - 238
Am - 242m
Cm - 242
Total
Sources of 225Ra, g atoms
U - 233
Np - 237
Am - 241
Pu - 241
Cm - 245
Total
5.75x10"'
5.47
2.30x10-'
6.22
1.02x10'
3.01
1.02x10'
8.06xlO-3
1.33x101
1.23
3.73
O6~
6.71xlQ-2
1.26x10-'
4.92X10-2
5.50x10- '
7.92x10- '
—
8.61x10'
5.47
5.15x10-'
2.30x10-'
9.23x101
U-fueled
PWR
discharge
fuel °J
1.15xIO?
5.47
2. -30x10-'
1.21x10?
1.02x10'
6.02xlO?
1.02x10'
8.06x10''
6.12x10 ''-
2.46xl02
3.73
2.50x10^
1.34x10'
2.52x10'
4.92xlO-;
5.50xlO-!
3.92x10'
—
8.61x10'
5.47
l.lSxlO2
2.30x10-'
2.07x10?
U-Pu- fueled "\
PWR
self-generated
Pu recycle a»D/
1.51
2.49x10'
7.07x10- '
2.71x101
8.97x10'
4.29
8.97x101
3.50x10-?
9.40x101
2.50
6.39x10'
6.64x101
5.68xlO-2
3.38x10-'
3. 28x1 O-2
2.95
3.67
__
6.37x10'
2.49x10'
1.51
7.07x10-'
9.08x101
35U-Th-fueled
PWR
U recycle £/
1.99
1.54x10-'
1. 17xlO-3
2.15
1.30x10-'
6.05
1.30X10-1
6.18
1.99
2.53xlO"2
2.02
5.05
3.36x10'
1.22xlO-3
4.95xlO-3
3.87x10'
9.58
1.17xl02
1.54x10-'
1.99
1.17xlO-J
1.29xl02
Pu-Th- fueled
PWR
U+Pu recycle v
4.42
l.lOxlO2
5.51
1.20xl02
1.37xl02
5.21
1.37xl02
—
1.42x10?
4.42
4.91x10'
5.35xlOJ
3.68
1.79x10-'
3.46
3.06
1.04x10'
8.96
1.33x10'
l.lOxlO2
4.42
5.51
1.42x10?
z»U-Th-fuele
HTSR
U recycle
2.21
7.97xlO"2
—
2.29
6.38x10-'
4.94
6.38x10-'
5.35x10-*
5.58
2.31
2.86x10-'
2.60
3.30
2.36x10"
1. 20xlO-3
2.02x10-?
2.69x101
4.06
4.60x10'
7.97x10-?
2.21
&• 24x10'
U-Pu- fueled
LMFBR V
9.62x10-'
1.68x10'
—
1.73x10'
7.90
3.05x10'
7.90
2.57x10-?
3.84x10"
1.03x10'
0.520
1 .08x101
2.88x10-3
2. 78x10- ?
2.94x10-'
4.68x10-'
7.93x10-'
2.16x10'
1.69x10'
9.62x10-'
3.94x10'
— High-level reprocessing wastes, 0.53J of U and Pu in discharge fuel appear in wastes. All other actinides in discharge fuel
appear in wastes. For equilibrium fuel cycles.
-' For PWR with 3.3X U fuel, E = 33 Mw day/kg, n = 0.325 (P2).
-^ High-level reprocessing wastes, 0.5% of U in first cycle 235U,and in bred Cl and 0.52 of Th appear in wastes. All other
actinides in discharge fuel appear in wastes.
-1 Source which contribute to the 22&Ra peak of -v-190,000 yr. 238U and 2u2Pu are not included.
-------
9. Generation of ltfC, 3H, and other Radionuclides
9.1 Carbon-14
Carbon-14 is an activation product of potential environmental
importance in the nuclear fuel cycle because of its long half life
of 5,730 yr and because it easily appears in volatile form, such as
C02. Most of the 1J*C formed in reactors results from the (n,p)
reaction with 14N:
in —, iifC + IH ,g ,v
0 61 V '
The lkH, which constitutes 99.6% of natural nitrogen, is present as
residual nitrogen impurity in oxide fuel of water reactors and fast-
breeder reactors, as air dissolved in the coolant of water-cobled
reactors, and as residual nitrogen in the graphite of high-temperature
gas-cooled reactors. The 14N activation cross section for 2200 m/sec
neutrons is 1.85 barns.
Carbon-14 also results from the (n,Y) reaction on 170, which is
present as 0.03% of natural oxygen, with a 2200 m/sec cross section
of 0.235 barns:
i70 + in — > !JC + jHe (9.2)
In graphite-moderated reactors another source of 14C is the (n,y)
reaction with 13C, which is present as 1.108% of the natural carbon
in graphite:
i63C + oin ->^C + °Y (9-3)
However, the 2200 m/sec cross section is only about 0.9 millibarns.
Additional but less important reactions are:
with a 2200 m/sec cross section of 2.4 x 10"7 barns, and
ISO + in -^C + fHe (9-5)
9-1
-------
The activity (Nx)~ of ltfC produced in a reactor can be estimated
by assuming irradiation in a constant neutron flux for a period TR.
Because of the long half life of 11+C, the approximation *CTR«1 leads to
(NX)C = XCTR I N^o (9.6)
i
where N. = number of atoms of species i_ producing ll*C by neutron reactions
a. = cross section for species i_ to produce ll*C
X- = radioactive decay constant for ll*C.
Carbon-14 produced in water coolant is important because of its
possible environmental release at the reactor site. If 1/+C forms
carbon dioxide or a hydrocarbon such as CH , and if no processes are
provided to recover the gaseous 11+C, the coolant-produced lkC will be
discharged along with the non-condensable gases removed by the main
condenser air ejector in a boiling water reactor and through the gaseous
waste disposal system for a pressurized water reactor.
We consider here the production of ll*C by reactions (9.1) and (9.2)
in the reactor coolant, which requires estimates of the inventories of
170 and dissolved nitrogen in the coolant within the reactor core. For
the 1000 Mwe PWR with an in-core water inventory of 13,400 kg, an
effective 170(n,a) thermal cross section of 0.149 barns, and an
average thermal neutron flux of 3.5 x 1013n/cm2sec, the 1UC production
from 170 is estimated to be 2.2 Ci/yr.
To obtain the 14C from dissolved nitrogen in the coolant, a dissolved
nitrogen concentration of one part per million (by weight) is assumed,
with an effective ll4N(n,p) cross section of 1.17 barns, resulting in a
yearly production of 0.061 Ci of l C. The total yearly production of l"C
in the PWR coolant is then about 2.3 Ci/yr, which is the source term for
possible environmental release at the reactor site. A 1000 Mwe boiling
water reactor would contain about 33,000 kg of water in the core under
operating conditions. Assuming the same values of neutron flux and cross
sections, the yearly production of 5.6 curies of 1IfC in the BWR coolant
is estimated.
The 11+C produced by 170(n ,a) in U02 fuel, calculated as the yearly
production per metric ton (Mg) of uranium originally in the make-up fuel,
is
106x6.02x1023 atoms U 2x3.74x10-'* atoms 170 c ... ,n 25 ?
- 238 -- M-U x - - x 6.47xlQ-25 cm2 x
3.5xl013 - — x - - x — - x 0.8 = 2. 54x1 0-2 Ci/yr Mg-U
cm2sec 3.7xl010dis/sec 5730yr
9-2
-------
For the ^N source in the fuel, it is assumed that the nitrogen
impurity is present in DO at a weight ratio of 25 ppm, although
nitrogen contents from 1 to 100 ppm have been reported (K?\ TH
yearly production per metric ton of U is
106 ^iS-FT x i * 25x10-6 9ram N . 6.02xl023 atoms
Mg U 238 gram U gram U02 x 14 grams -
0.996 atoms .ltfN , ,, ln-2l» 2 13 i r,
- x 1.17x10 cm x 3.5x10 — - - x _ —
atom N cm2sec a.yxlO^dls/sec
x 0.8 = 0.130 Ci/yr Mg U
The total amount of ll+C produced yearly in the fuel is then 0.153 Ci
per metric ton of uranium.
To obtain the 14C in the discharge fuel, we use the fuel life of
three calendar years for the reference pressurized-water reactor. Since
there is negligible decay of the 1'*C during this 3-yr period, the con-
centration in the discharge fuel is
3 x 0.155 = 0.466 Ci/Mg
The quantity of llfC in the total fuel discharged yearly,'which initially
contained 27.2 Mg of uranium, is:
0.466 x 27.2 = 12.7 Ci/yr.
In a pressurized water reactor operating with plutonium recycle
the thermal neutron flux is lower than for uranium fueling because of
the higher fission cross section for plutonium. As a result, less
ll*C is produced by thermal-neutron activation within the fuel, as shown
in Table 9.1.
9-3
-------
TABLE 9.1 1L*C in Discharge Fuel (1000 Mwe reactors,
80% capacity factor)
PWR PWR HT6R LMFBR
U U and recycled 235U, Th, and U and recycled
(3.3% 235U) U + Pu recycled U Pu
l*C, Ci/yr 12.7 6.7 IZO^-7 3.3
24i>/
-Calculated for 30 ppm N, in HTGR graphite
h /
-Calculated for 1 ppm N2 in HTGR graphite
9-4
-------
Fast-breeder oxide fuel is also assumed to contain 25 ppm
of residual mtrogen(KZ). Typical average fast-spectrum cross
sections are 0.135 millibarns for ^0(n,Y) and 14 millibarns for
wN(n,p) within the reactor core (C3). For an average fast-spectrum
core flux (C3) of 3.8xl015n/cm2sec, and for the parameters of a near-
term Pu02LMFBR (P2), the yearly production of a"C for a 1000 Mwe fast
breeder is estimated to be 3.3 Ci/yr. Relatively little JltC is pro-
duced in the blanket fuel because of the Tower neutron flux there.
The fuel of the high-temperature gas-cooled reactor (HTGR)
consists of uranium and thorium particles, as oxides and carbides,
distributed through a graphite matrix. The important ^C-producing
reactions in this fuel are 11+N(n,p) and 13C(n,Y). Residual nitrogen
is assumed to be present in graphite at a weight ratio of 30 ppm (B7).
In the thermal-neutron energy spectrum of an HTGR the effective
activation cross sections (B7)are 0.683 barns for 11+N and
3.3x10 ** barns for 13C. For the average thermal-neutron flux of
1.2xl0ll+n/cm2sec and a 4-yr fuel life, the estimated concentration
of 14C in the discharged graphite fuel is calculated from Eq. (9.6),
with the result:
Curies of 14C per kg
of graphite in discharge
Source fuel
i"N(n,p), 30ppm N 1.10 x 10"3
13C(n,Y) 2.29 x IP""
Total 1.33xlO"3
The fuel discharged yearly from the 1000 Mwe HTGR reactor of Fig. 5.1
contains 7 95 Mg of heavy metal and 90.5 Mg of graphite The yearly
production of 1?C by this reactor is then estimated to be
1.33x10"3x 90,500 = 120 Ci/yr
In other HTGR calculations 1 pom of H2 1n the graphUe 1s assumed (H4i
resulting in an estimated yearly production of 24 Ci/yr for a luuu ne
plant.
in oxygen
9-5
-------
gas, which contains the 14C and all of the normal carbon from the
graphite, is to be recovered to avoid release of 11+C to the
environment.*
The greater ll*C production in the HTGR, whether a factor of
2 or 10 times greater than in the oxide-fueled water reactors, is
probably not the main issue in comparing the management of llfC wastes
in the HTGR cycle with 14C management in LWR and CANDU cycles utilizing
urania or thoria. When urania or thoria are reprocessed the ll*C
released in fuel dissolution is diluted by normal carbon, in the form
of C02 in the dissolver off-gas. An isotopic concentration of 11+C
in CCL from dissolving oxide fuel of 130 to 650 ppm is estimated
(D3). This is relatively concentrated when comoared with the 11+C
content of C02 released in HTGR fuel reprocessing, where very large
quantities of normal carbon (90.5 Mg/yr) form C02 when the graphite
fuel is incinerated. For 30 ppm N in the graphite, the resulting
isotopic concentration of !t*C in carbon is 0.3 ppm, and it decreases
to 0.06 ppm for 1 ppm N in the graphite. This large volume of C02 ,
containing relatively small concentrations of radioactive gases, creates
a challenging problem for fuel reprocessing development. Because
the C02 interferes with the processes normally used to concentrate and
remove 85Kr from air streams, a new krypton-removal process is under
development for HTGR fuel reprocessing.
In the HTGR reprocessing, the incinerator gases contain con-
siderable carbon monoxide, so the filtered gas is first passed over
a catalyst to oxidize CO to C02. Also, that portion of the tritium
which may be in the form of HT is oxidized to HTO. Elemental radio-
dine is removed from the C02 by adsorption on a bed of lead zeolite,
followed by a bed of silver zeolite for final elemental iodine clean-
up and to remove methyl iodide. Tritiated water is removed on
molecular sieves. Because of the low concentration of HTO in the
C02 gas, it may be necessary to inject steam or water vapor upstream
of the adsorbent bed as a carrier for HTO removal. After removal of
220Rn and 85|
-------
of CaQ) containing ltfC may have to hp v-pac^ -
to al phi-contaminated wastes I threshold of iJ to" Ton™
o
per gram for alpha-contaminated wastes an t pra t al es0
monitoring at this level would evidently dictate ?hat e ent a?lv
all of the reprocessing wastes must be treated as TRU wastes to
be disposed of ultimately in a geologic repository. The alpha
contamination of the CaOO is unknown. Even if the CaCO s not
a pha contaminated, the l^gic of the 10 to 100. na cu e/3gr L
threshold may apply as well to these i*c wastes, if this were to
result In a requirement that the HTGR-produced CaCO, be emplaced
in the federal geologic repository, along with TRU wastes, a
substantial penalty could accrue to the HTGR fuel cycle.
Croff(C4) has analyzed the cost of various alternative means of
managing the CaC03 produced in HTGR reprocessing. Burial in a
geologic repository was found to be the most expensive of the al-
ternatives considered, with an estimated cost in constant 1975 dollars
of $280 per kilogram of heavy metal (Th + U) reprocessed. Because
of the high burnup of the HTGR cycle, the economics of this fuel
cycle are less affected by high unit costs of fuel cycle opera-
tion than in the case of the thoria fuel cycles. However, Croff's
estimate for CaC03 disposal is over five- fold greater than the
total cost, in the same constant dollars, estimated for off-site
disposal of all wastes from reprocessing urania fuel (U4).
Assuming generally similar wastes and waste costs from thoria
fuel as from urania fuel, the additional costs of disposing of
the large quantities of ^C-contaminated CaC03 may impose a
significant economic penality on the HTGR fuel cycle.
The issue of ll*C and how it must be disposed of also illustrates
the problem of developing adequate and meaningful criteria for
long-term waste management. As compared to the total actinides in
reprocessing wastes, 14C contributes little to the total activity.
Its contribution to total ingestion toxicity is even lower, because
its RCG is several orders of magnitude less than that of the actinides.
When using the calculated waste toxicity as a criterion of hazards,
ll*C would be considered to be relatively unimportant.
However, an evaluation of the hazards from waste disposal must also
take into account the mechanisms and probabilities of the radionuclides
reaching the biosphere. As an example of such an approach, Burkholder
(88, B9) has calculated the migration of radionuclides through a geo-
logic medium. He assumed that all radionuclides in the geologic repos-
itory are leached into ground water at the same rate. The different
equilibria between the various diffusing species and the soil through
which they migrate were taken into account, and a constant linear
velocity of ground water was assumed. The calculated amounts of
radionuclides which appear at various distances and at various times
from the position of emplacement indicate that, for the desert-son
9-7
-------
properties and other conditions assumed by Burkholder, ll*C delivers
a greater whole body and organ dose than any of the other radio-
nuclides in the reprocessing wastes, which reflects the relatively
high mobility of carbon compounds in such geologic media. Burk-
holder's ground-water model is highly simplified, and the para-
meters assumed in his calculations may not be aopropriate for
waste forms and geologic media that are finally selected. However,
his calculations do illustrate the kind of methodology that should
be develooed, and they suggest that it is premature to draw con-
clusions on the importance of geologic isolation of lkC on the basis
of the calculated toxicity index.
From Croff's analyses (C4), it appears that the importance of
waste mananement to the HTGR fuel cycle is already recognized.
This problem is also being addressed in other countries. In Japan
research is underway (F13) to incinerate the graphite fuel prisms
with C02 rather than oxygen. The resulting CO is then to be cat-
alyticaily decomposed into elemental carbon and C02, the C02 is then
recycled to oxidize more graphite. In this way a pure carbon waste
is produced, which is of smaller volume and lower solubility than
CaCO.v The method of disposing of this ^C-contaminated waste has
not been determined.
Definition of the requirements for 14C waste disposal appears
to be an important step which may significantly affect the choice
of the HTGR fuel cycle.
9.2 Tritium (3H)
Tritium formed in or released to the reactor coolant is a potential
environmental contaminant of the reactor site, and tritium remaining
with the discharge fuel is a potential contaminant of the reprocessing
plant. In light water reactors the dominant source of tritium is from
ternary fission. For a 1000 Mwe pressurized water reactor fission-
product tritium is formed at the rate of l.SSxlO14 Ci/yr for uranium
fueling and 2.47x10'* Ci/yr with self-generated plutonium recycle. The
estimated rate of formation of tritium in the reactor coolant is shown
in Table 9.2. All of this coolant tritium is released to the environ-
ment at the reactor site, largely in the form of highly diluted HTO in
liquid effluents.
In the HTGR the principal non-fission source of tritium is from
6Li(n,a) in lithium contaminants in the graphite and core matrix.
Lithium concentrations of 0.01 to 1.2 ppm in HTGR graphite have been
reported (G2, H4).
9-8
-------
TABLE 9.2. Estimated Tritium Production in the Coolant of a 1000
Pressurized Water Reactor.
Tritium production
Source Ci/yr
2H(n,Y)
360
6Li(n,a) 34
7Li(n,a)T _4
Total from activation reactions 400
Fission-product tritium- 188
TOTAL 588
-/ Assumes fission-product tritium diffusing through fuel cladding or
escaping through pin-hole cladding failures is equivalent to release
of fission-product tritium from 1% of the fuel.
9-9
-------
The tritium thus formed evidently diffuses to the coolant,
so we shall estimate the average yearly production of 3H in the
coolant due to 6Li(n,a). At much lower concentrations the lithium
is exposed homogeneously, to the neutron flux. Because of its
large thermal-neutron cross section, 6Li is significantly depleted
during the typical fuel irradiation time of 4 years. The average
yearly production of 3H from this reaction is then given by
where
^T
'6
\
FR
1-e
1-e
-ATTR
XTTR
= radioactive decay constant for 3H
= initial number of atoms of 6Li
= (N,a) cross section of 6Li
= irradiation time of discharge fuel
(9.7)
For a core inventory of 362 Mg of C for a 1000 Mwe HTGR, and
neglecting production of 3H in the graphite reflector, we obtain
3H from 6L(n,a) = 232 Ci/yr. for 0.01 ppm Li
= 2.79 x TO4 Ci/yr. for 1.2 ppm Li
From calculations by Gainey (G2) of the 10B(n,T) activation due to
boron in HTGR control absorbers and burnable poisons, an additional 1250
Ci/yr of tritium is formed and diffuses to the coolant. Also reaching
the HTGR coolant is about 0.5% of the fission-product tritium formed
within the fuel particles, tritium formed in boron control absorber,
and tritium formed by (n,p) reactions with 3He impurities in the coolant.
Because of the relatively large quantities of tritium thus formed,
it is necessary to remove the tritium by reacting it with hot titan-
ium in the continuous coolant clean-up system.
The small amounts (1.7xlO~5%) of 3He present in underground sources
of natural helium used for the HTGR coolant produces tritium by the
reaction:
|He
(9.8)
with a cross section of 5326 barns for 2200 m/sec neutrons and an
effective cross section of 2800 barns at the HTGR operating temperature.
9-10
-------
For an inventory of natural helium of 618 kg in the core of
a 1000 Mwe HTGR (B12), 3H is initially formed at the rate of about
8,020 Ci/yr. However, because of its large cross section, 3H is
rapidly depleted by neutron absorption. It is replaced by fresh
helium introduced to make up for coolant leakage. If a fraction
fue of the coolant leaks from the coolant system per unit time,
the steady state concentration XH . of 3He within the reactor
coolant can be calculated by J
NHe fHe 4-3 = C XHe-3 *°He-3 + NJe XHe-3 fHe <9
where N! = total inventory of helium in the coolant system
NU = total inventory of helium within the reactor core
= atom fraction of 3He in natural helium (1.7xlO~7)
le-3
Solving for XHe_3, we obtain
(9.10)
XHe-3
= XHe-3
NHe *aHe
' 4 fHe
From HTGR design data, it is estimated (B12) that
NR
:"§_ = 0.09
NHe
fHe = 0.015/yr
9-n
-------
For an effective aH - = 2800 barns, and for $ = 1.2xl01Vcm2sec,
we obtain
XHe-3 =2-63x10-9
The resulting steady-state rate of production of tritium in the
coolant from 3He(n,p) is 124 Ci/yr.
In the CANDU heavy-water reactor the dominant source of tritium
is the deuterium activation reaction. Data given for the Douglas Point
Nuclear Power Station (C6) provide a basis for estimating the rate
of production of tritium in the heavy water moderator and coolant:
electrical power = 203 Mwe
inventory of D20 coolant in reactor core = 2.45xl03g
average thermal neutron flux in coolant = 6.10xl013/cm2sec
inventory of D20 moderator in reactor core = 7.18xl07g
average thermal neutron flux in moderator = 1 .Olxl0ltf/cm2sec
average 2H(n,y) cross section = 4.45x]0-'t barns
The rate of production of 3H in the D20 is then:
(2.45xl03 x 6.10xl013 + 7.18xl07 x I.
6.02x1023 x 2 atoms2H an 2
s\ ^irt f\ f\ r\ r\ ** ^ ^ o« • **
cm2sec
x 4.45xlO"28cm2
Ci
20.02g D90
12.Syr A 3.7xlO^Ysec
x 0.8 = 2.42x10 Ci/yr
For a 1000 Mwe CANDU power plant with the same reactor lattice and
with the same ratio of D20 in core inventory to uranium inventory as
in the Douglas Point Reactor, the yearly production of tritium in
the heavy water would be
1000
203
x 2.42xl05 = 1.19xl06 Ci/yr
Because of this large rate of tritium generation it is necessary
to operate a small isotope-separation unit to prevent the build-up
of large concentrations of tritium in the heavy water. The losses of
heavy water are kept small enough so that only a very small fraction
of the tritium is released to the environment. The yearly release
of tritium reported for the Douglas Point Station is typically about
4000 Ci/yr, which is about 0.2% of the allowable release (D4)-
The amounts of tritium produced annually by these different
reactors are summarized in Table 9.3.
9-12
-------
TABLE 9.3 Summary of Tritium Production in Reactors
Reactor Type PWR
Fuel U
Fission-product 3H, Ci/yr l.SSxlO1*
3H in coolant, Ci/yr 5.88xl02 -
PWR
U and self-
generated Pu
recycle
2.47xl04
/ 6.47xl02
CANDU
HWR
U
l.SSxlO1*
1.19xl06 -1
HTGR
235 U, Th,
and recycled
U
9.59xl03
1.65xl03 y
2. 93x1 01* -1
-1
See Table 9.2
- D20 coolant + D20 moderator
- 0.01 ppm Li in C, 0.5% release of fission- product tritium
- 1.2 ppm Li in C, 0.5% release of fission-product tritium
9-13
-------
9.3 Sulfur- 35. Phosphorous-33, and Chlorine- 36 in HTGR Fuel
The graphite fuel blocks of the HTGR reactor contain sulfur con
taminant, which originates from the pitch used to form the fuel-rod
matrix material. Neutron activation of the 4.22% 34$ in natural
sulfur results in 88- day 35S, according to the reaction:
for which the 2200 m/sec cross section is 0.24 barns. Assuming
that sulfur is present at 193 ppm in the HTGR fuel(H4), it is
estimated that 215 Ci of 35S are present in the fuel discharged yearly
from a 1000 fV/e HTGR reactor, after 150 days of storage. In the
HTGR reprocessing the stable and radioactive sulfur will volatilize
to follow the carbon dioxide from graphite incineration. The
radioactive sulfur is a potential environmental contaminant that
must be recovered. The amount of 35S activity is greater than that
of 1/+C, and the inhalation RCG is over an order of magnitude lower
for 35S. The stable sulfur may interfere chemically with some of
the recovery processes in the off-gas system.
Natural sulfur also contains 0.76% 33S, which undergoes (n,p)
reactions to form 25-day 33P, according to
fg 10)
\v-it-i
with a 2200 m/sec cross section of 0.14 barns. The estimated activity
of 33P in the fuel discharged annually from a 1000 Mwe HTGR, after 150
days of storage, is 1.1 Ci .
Another volatile radionuclide formed in HTGR fuel is 3.1xl05-yr
36C1 , formed by neutron activation of chlorine contaminant in the
fuel, according to the reaction:
(9.13)
Natural chlorine contains 75.77% 35C1, for which the 2200 m/sec
activation cross section is 43 barns. Assuming 3 ppm chlorine in
the fabricated HTGR fueHH4), the estimated yearly production of
36C1 from a 1000 Mwe reactor is 1.02 Ci .
•
These additional radionuclides volatilized in HTGR fuel reprocess-
ing are summarized in Table 9.4.
9-14
-------
TABLE 9.4. Additional Volatile Radionuclides in HTGR Discharge Fuel
(1000 Mwe 235U-Th-fueled HTGR, 80% capacity factor,
150 days storage)
Ci/yr
35$ 215
33R 1.1
1.02
9-15
-------
9.4 Non-Volatile Radionuclides Activated in Fuel Element Structure
Fuel elements discharged from pressurized water reactors also
contain radionuclides formed by neutron activation in the Zircaloy
cladding, stainless steel end fittings, and Inconel spacers. A
typical three-year irradiation of the metallic structure produces
the radionuclides listed in Table 9.5, calculated for fuel elements
discharged from a light-water reactor and stored for 150 days (BIO).
Neutron capture in stable 9l*Zr forms 65-day 95Zr and its decay
daughter, 35-day 95Nb. The radioactivity produced is large, but
it is still smaller than the radioactivity of these two nuclides
formed as fission products. Other large contributors to the cladding
radioactivity are 60Co, resulting from neutron capture in stable
59Co, and 51Cr, 55Fe, 58Co, and 68Ni.
After 10 years of decay there is still appreciable radioactivity
remaining, so irradiated cladding must be treated as a long-lived
radioactive waste. The only species which persist after about a
thousand years of decay are 1.5xl06-yr 93Zr and 2.12xl05-yr 99Tc.
The activity of 93Zr in irradiated cladding is about the same as
the activity of fission-product 93Zr, but the activity of 99Tc in
cladding is about 1000 times less than the activity of fission-product
"Tc.
The fast-breeder fuel cladding and structure, typically of 316
stainless steel, result in the radionuclides listed in Table 9.5 (BIO).
Since the structure is entirely an austenitic alloy, the most radio-
active nuclides are 5l*Mn, 55Fe, and 60Co.
The HTGR fuel contains no metallic structure, but impurities
in the graphite fuel blocks result in the production of relatively
small amounts of radioactive cobalt and nickel, as listed in Table
9.5 (H4). The total activity from metallic contaminants in HTGR fuel
is considerably lower than that in the fuels from light-water and
breeder reactors.
9-16
-------
TABLE 9.5
Nonvolatile Radionuclides In Discharge Fuel Prom Neutron Activation
(1000 Mwe reactors, 80% capacity factor)
Activity in discharge fuel, Ci/yr
Reactor type-'
Fuel
Beryllium 10
Sodium 22
Phosphorus 32
33
Calcium 45
Scandium 46
Vanadium 49
Chromium 51
Manganese 54
Iron 55
59
Cobalt 58
60
Nickel 59
63
Strontium 89
Yttrium 91
Zirconium 93
95
Niobium 92m
93m
95
Molybdenum 93
Technetium 99
Tin 117m
119m
121m
123
Antimony 124
125
Half Life
2.5xl06yr
2.60yr
14.3day
25 day
165 day
83.9day
330 day
27.8day
303 day
2.6yr
45 day
71.3day
5.26yr
8xl04yr
92 yr
52 day
58.8day
1.5xl06yr
65 day
10.16day
35 day
>100 yr
2.12xl05yr
14.0day
250 day
76 yr
125 day
60 day
2.7yr
PWR HTGR
U 235U,Th, and
(3.3% 0) recycled 0
1.20X10"1
-2
4.61x10
3.37X101
1.91X104
4.79X103
4.89X104
6.17x10
5.92x10, 0.244
1.66x10 4.46x10
l.OSxlO2 1.72 ,
1.56x10 2.28x10
1.41X102
4.69X102
2.81
1.59x10
2.90X10"1
2.96x10
5.45X10"1
S.Slxlo"1
1.96X102
4.31x10
9.16
5.30
2.28x10?:
1.10x10
LMFBR
U and recycled
Pu
5.16*
23.7
3.16
7.04X10"1
2.03xl04
1.74X106
1.30x10*
1.47x10
2.24x10*
3.22x10
7.46x10*
2.37x10
2.09X10*1
4.86
4.88x10
7.46X101
7.25
Tellurium 125m
TOTAL
58 day
4.97x10
7.72x10
2.61x10
5.33x10
a/ PWR = pressurized water reactor
HTGR - high-temperatue gas-cooled reactor
LMFBR = liquid-metal-cooled fast breeder reactor
Data are calculated for 150 days after discharge for PWR and HTGR,
60 days after discharge for LMFBR. 9-17
-------
9-5 232U in Uranium Recovered From Irradiated Thorium
The results of several different calculations of the concen-
tration of 23ZU in uranium recovered from irradiated thorium and re-
cycled uranium are summarized in Table 9.6. The data of Shapiro (SI)
for a pressurized-water reactor are the same as the 232U concentra-
tions appearing in the recycled uranium of Table 3.5 and 3.7,
wherein all uranium in the discharged fuel is assumed to be re-
cycled, except for process losses. In the case of 235U make-up
the recycled uranium includes uranium bred from thorium as well
as residual uranium from the 235U make-up. These data are quoted
for the fifth generation of irradiation, i.e. the build-up of
232U has been followed through each generation consisting of a
full irradiation exposure followed by reprocessing, uranium re-
covery, and fabrication of that recycled uranium with additional
thorium for the next generation of irradiation.
232U concentrations calculated by Arthur and quoted by
Rainey (Rl) for a PWR fueled with thorium and 233U make-up
are considerably higher than those calculated by Shapiro (SI);
possibly because of the high initial 232U concentration (1300 ppm)
assumed by Arthur for the make-up 233U. The Arthur-Rainey
results indicate a 15% increase in the 232U concentration if the
initial thorium contains 100 ppm 230Th.
Arthur's (A2) calculations for a PWR fueled with thorium
and denatured uranium indicate far less isotopic concentration
of 232U in this fuel cycle, evidently because of the dilution by
the denaturing 238U (see Chapter 7).
Mann and Schenter's (M4) calculations for an oxide-fueled
LMFBRwith 233U-232Th core fuel indicate equilibrium 232U concen-
trations near the 232U concentrations predicted by Shapiro for
the thorium-fueled PWR's.
9-18
-------
TABLE 9.6. Summary of Calculations of 232U in Recycled Uranium Recovered from Irradiated Thorium
Reference
Reactor Fuel
assumed 230Th in
make-up Th,
BE!"
calculated 232U in
recycled U,
ppm
10
10
Shapiro (SI)
Shapiro (SI)
Rainey-Arthur (Rl)
Arthur (A2)
General Atomic (H4)
Mann and Schenter (M4)
PWR thorium + 93% 235U make-up,
all U is recycled,
near-equilibrium (5th generation)
PWR thorium + Pu,
all U and Pu are recycled,
near-equilibrium (5th generation)
PWR thorium + 233U make-up,
all U is recycled,
near-equilibrium (5th generation)
PWR thorium + denatured U,
235U make-up,
near-equilibrium (5th generation)
thorium + denatured U,
233U make-up,
near-equilibrium (5th-generation)
HTGR thorium + 93% 235U make-up
recycled bred uranium,
near-equilibrium (2nd generation)
LMFBR 10% 233U + 90% 232Th in core,
(oxide fuel) 2-year irradiation,
equilibrium
100% Th blanket,
3-year irradiation
0
100
0
85
0
85
100
2600
2800
4000
4600
260
316
512
563
742
2760
86
-------
10. Summary and Conclusions
The present commercial light-water reactors and the CANDU heavy-
water reactors can be adapted to thorium fueling with very little modi-
fication in reactor design. For the thorium fueling in these reactors
to be useful, a closed fuel cycle is required, involving the repro-
cessing of discharge fuel and recovery and recycle of fissile material.
The rep recessing-re fabrication technology for urania-thoria or urania-
plutonia-thoria fuel is basically similar to that for urania-plutonia
fuel, but more development and scale-up experience is required before
the closed fuel cycle for thoria systems can be implemented.
The near-commerical HTGR is already designed for 235U-Th fueling.
However, reprocessing and refabrication operations for HTGR fuel differ in
many respects from the present technological base established for
urania-plutonia fuels. Considerably more development, beginning at
the pilot-plant scale, is required.
Fueling LWR's and HTGR's with thorium and with plutonium recovered
from uranium fuel discharged from LWR's is a logical way to introduce
thorium fueling, but it achieves no better savings in uranium resources
than recycling this plutonium as mixed-oxide urania-plutonia fuel in
light-water-reactors. About 20% further savings in the reactor-lifetime
uranium ore requirement is possible if the LWR or HTGR is fueled initially
with thorium and 93% make-up 235U, with recovered uranium to.be recycled.
This ore saving is calculated for first-gene ration reactors that must
begin operating with no equilibrium fissile inventory in the reactor
and fuel cycle. The time to reach equilibrium is relatively long in the
uranium-thorium cycle.
The CANDU heavy-water reactor requires 40% less uranium ore when
fueled with natural uranium and recycled self-gene rated plutonium than
does the LWR with self-generated plutonium recycle. For a CANDU reactor
started with natural uranium and converted to plutonium-thorium fuel-
ing as plutonium is recycled, the lifetime uranium ore requirement is
less than half that of a uranium-fueled LWR with self-generated plutonium
recycle. If started initially with 93% 235U-Th, the CANDU lifetime uranium
requirement is reduced to 39% of that of the uranium-fueled LWR with
self-generated plutonium recycle. CANDU fuel elements must be modified
to accomodate the higher burnups associated with these fissile-recycle
fuel cycles. If started with natural uranium and converted to thorium
fueling as self-generated plutonium is recycled, the CANDU reactor can
reach self sustaining breeding, with a total uranium ore requirement
within about 23 years, with a total uranium ore requirement 45% of the
lifetime requirement of the uranium-fueled LWR with self-generated
plutonium recycle. No additional uranium ore would be required for
subsequent replacement thorium breeders.
The larger uranium ore savings possible with the CANDU reactor are
a consequence of the relatively large conversion ratio of this reactor,
which is 0.75 with uranium fueling, an average of 0.92 with the modes of
high-burning thorium fueling analyzed in the present study, and 1.0 for
low burnup thorium fueling with uranium recycle. Other studies indicate
that with some lattice modification the CANDU conversion ratio can be
further increased.
10-1
-------
Improvements in conversion ratio of the PWR and HTGR, accompanied
by further reduction in lifetime uranium ore requirements, are possible,
but they appear to involve considerable design modification to achieve
even the present conversion ratios of the CANDU reactor.
Given a reactor industry already based upon LWR's, the most
direct and resource-effective approach to conserve uranium resources
is to use plutonium from water reactors to start piutonium-uraniurn-
fueled fast-breeders. Given a stockpile of depleted uranium as an
already mined resource sufficient to fuel fast breeders for thousands
of years at the end-of-century energy demand, the natural uranium re-
source required for the fast breeder is that attributable to supplying
the start-up plutonium for the first-generation breeder. This ore for
breeder startup represents a 32% increase in the lifetime uranium ore
requirement for the light-water reactors producing the plutonium for
the replacement breeders, assuming that these light-water reactors
would otherwise operate with self-generated plutonium recycle. Plutonium
start-up is the most resource-effective start-up approach for fast
breeders. Thorium fueling with fast breeders offers no resource ad-
vantage, in the absence of special constraints that may be introduced
by safeguards considerations.
Safeguards considerations have led to concerns about the recovery
and utilization of plutonium in the power-reactor fuel cycle. The logic
of the possible use of plutonium for nuclear explosives applies equally
well to 93% 235U. Therefore, until these concerns are resolved, cycles
involving plutonium recycle or implementation of the Pu-Th or 93% 235U-
Th fuel-cycle alternatives might require that the reactors and repro-
cessing-refabrication operations involving these fuels be under special
security control., such as location in specially controlled "international"
centers.
These safeguards concerns have suggested the possibility that de-
natured uranium, i.e., uranium containing a low enough fissile concentra-
tion to be unsuitable for nuclear explosives, is sufficiently self-pro-
tected that reactors operating with such fuel can be safely dispersed as
national reactors. The fuel discharged from these dispersed reactors
W11' contain Plutonium, but the plutonium in that form is assumed to be
sufficiently self-protected by the intense radioactivity of the fission
products. The discharge fuel would be shipped to the international center
for reprocessing, and recovered plutonium would be consumed in plutonium-
burning reactors colocated at the international center. From these analyses
herein it is concluded that:
10-2
-------
(a) Present 3% 235u_238u LWR fuels and natural uranium CANDU
fuels are suitably denatured for such dispersed reactors. To obtain
uranium-resource benefits from plutonium utilization, the power of
piutoniurn-burner reactors at the international center must be an
appreciable fraction of the total power of the nuclear power system.
For LWR reactors, the ratio of international piutoniurn-burner power
to the power of dispersed natural reactors is 0.4.
(b) Plutonium production can be suppressed, and the necessary
relative power of the international pi utoni urn-burning reactors re-
duced, by fueling the national reactors with thorium and denatured
233U-235U-238U. Still using an international U-Pu-fueled PWR, fueling
national reactors with thorium and denatured 233y_235u_238u reduces the
ratio of Pu-burning-reactor power to disnersed-reactor power to 0.14 for
LWR dispersed reactors and to 0.07 for CANDU dispersed reactors. Further
reduction is possible with Pu-Th-fueled plutonium burners.
(c) Similar combinations are possible with plutonium-thorium-
fueled fast breeders, located at the international center, furnishing
233U for the dispersed national reactors. However, this breeder power
must be relatively large and the effective doubling time for breeder
fissile-inventory is considerably lengthened, thereby decreasing the
rate at which breeders can be introduced.
(d) Denatured 233U-238U, with a fissile concentration of about
12% 233U, is relatively vulnerable to non-commercial isotopic enrichment
to concentrations possible for explosives.
The total alpha activity of recycled plutonium in the uranium-
plutonium fuel cycles is considerably greater than the alpha activity
of recycled uranium in a uranium-thorium fuel cycle for the same reactor
capacity. The higher-energy gammas from 232U daughters accompanying recycled
233U may require greater-shielding in fuel fabrication than in uranium-
plutonium systems. The largest quantities of 232U are calculated for
uranium-thorium fuel in light-water lattices. Thorium recovered from
irradiated fuel must be stored for many years for 228Th decay before
it can be recycled; 3 to 17 years are estimated for the reference HTGR
U-Th fuel cycle.
The HTGR discharge fuel, whether from the U-Pu or U-Th
fuel cycle, will contain relatively large quantities of 14C diluted by
a large amount of non-radioactive carbon from graphite incineration.
The disposition of this long-lived solid waste is an environmental
issue which warrants further study. The relatively large production
of tritium in the CANDU heavy-water reactor is an environmental feature
of this reactor, whether fueled with uranium or thorium.
It is the choice of fuel cycle, rather than the choice of the
reactor, which has the greatest effect upon the long-term radioactivity
and ingestion radio-toxicity of the high-level radioactive wastes.
Differences in long-term radioactivity and toxicity are due more to
differences in actinide composition and production, rather than to
10-3
-------
differences in the yields of fission products. In 235U-Th fueling
relatively little americium and curium are formed, and relatively
little 239Pu and 21+0Pu appear later from the decay of americium and
curium. Consequently, the ingestion toxicity of U-Th fuel-cycle
wastes is relatively small during the period beginning at about
600 years after reprocessing, when 90Sr and other fission products have
decayed, until about 30,000 years. Relatively large quantities of
23!(U and 238Pu in the wastes from the U-Th fuel cycle result in a peak
in 22GRa radiotoxicity at 190,000 years of greater magnitude than in any
of the other reprocessing fuel cycles and comparable to that in un,re-
processed discharge uranium fuel.
With Pu-Th fueling relatively large quantities of Am, Cm, and Pu
appear in the wastes. The long-term radiotoxicity due to these actinides
in wastes is within an order of magnitude of the long-term radioactivity
of the same radionuclides in unreprocessed discharged uranium fuel.
The long-term ingestion radiotoxicity of thorium mill tailings is
less than that of uranium mill tailings in the 235U-Th near-equilibrium
fuel cycle, provided the natural thorium contains no contaminant 230Th.
If the natural thorium contains 100 ppm 230Th, the ingestion toxicity
of thorium mill tailings is increased to ten times that of the uranium
mill tailings.
10-4
-------
11. Acknowledgments
The authors express special appreciation for the guidance and
assistance provided by Mr. Bruce Mann, Chief of the Evaluation Branch,
Office of Radiation Programs, Las Vegas Facility, who was the EPA
Project Officer for this study.
The study was supported in part by the Energy/Environment
program of the EPA Office of Energy, Minerals, and Industry,
Assistant Administrator for Research and Development, and in part
by the University of California. Funding for computer time was
provided by the EPA Office of Radiation Programs.
This final report reflects the many useful and constructive
suggestions and comments resulting from reviews of an earlier draft
by several agencies and individuals, including:
DOE, Nuclear Power Development Division, Assistant
Director for Fuel Cycle, Office of Waste Management
EPA, Office of Radiation Programs, Technology Assess-
ment, Division staff
Dr. C.E. Till, Director Applied Physics Division, Argonne
National Laboratory
Oak Ridge National Laboratory Staff
Dr. Bal Raj Sehgal, Program Manager Nuclear Safety and Analysis
Department, Electric Power Research Institute
Dr. Norton Shapiro, and Dr. R. A. Matzie, C-E Power Systems,
Combustion Engineering, Inc.
Professor S. Banerjee, Department of Engineering Physics,
McMaster University, Ontario, Canada
Mr. Mitsuru Maeda, Japan Atomic Energy Research Institute,
Tokai, Japan. During his recent appointment as Research
Associate at the University of California, Mr. Maeda reviewed
many parts of the report and contributed directly to fuel-
cycle calculations for Chapter VII.
The many sources of outside information, both through published
work and by direct input from individuals, are identified and credited
in the list of references.
The authors are grateful for the help of Mrs. Sue Thur in preparing
the manuscript and to Mrs. Edith Boyd, EPA, for proofreading and eaitorfal
assistance.
The contents of the final report, including interpretations and
conclusions therein, remain the sole responsibility of the authors.
11-1
-------
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B8 H. C. Burkholder, M. 0. Cloninger, D. A. Baker, G. Jansen,
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1975.
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" Radioactive Wastes to be Generated by the U. S. Nuclear Power
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B12 A. Baxter, General Atomic, Private Communication, February, 1978.
Cl R. J. Cholister, W. A. Rodger, R. L. Frendberg, W. L. Godfrey,
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12-1
-------
C2 E. Critoph, S. Banerjee, F. W. Barclay, D. Hamel, M. S.
Milaram, J. I. Veeder, "Prospects for Self-Sufficient Equilibrium
Thorium Cycles in CANDU Reactors," AECL-550.1, 1976.
C3 A. G. Croff, Oak Ridge National Laboratory, Private Communication,
August, 1977.
C4 A. G. Croff, "An Evaluation of Options Relative to the Fixation
and Disposal of 11+C-Contaminated C02 as CaC03,"ORNL-TM-5171,
April, 1976.
C5 E. Critoph, "The Thorium Fuel Cycle in Water-Moderated Reactor
Systems," AECL-5705, May, 1977.
C6 B. L. Cohen, "High-Level Waste From Light-Water Reactors,"
Rev. Mod. Phvs. 49_, 1977.
C7 W.R. Cobb, W.J. bich, and D.E. Tivel, "Advanced Recycle .Methodology
Program System Documentation," EPRI CCM-3, Part II Chapter 5,
September, 1977.
Dl R. C. Dahlberg and L. H. Brooks, "Core Design Characteristics
and Fuel Cycle," Nucl. Eng. Inter. 640-646, August, 1974.
D2 W. Davis, Jr., "Carbon-14 Production in Nuclear Reactors," ORNL/
NUREG/TM-12, February, 1977.
D3 W. Davis, Jr., Oak Ridge National Laboratory, Private Communication,
November, 1977.
D4 T. S. Drolet, E. C. Choi, and J. A. Sovka, "CANDU Radioactive
Waste Processing and Storage," Trans. ANS 22, 354, November, 1975.
D5 F. E. Driggers, "Reference Thorium Fuel Cycle," E. I. DuPont De
Nemours & Co., Savannah River Laboratory, DPST-TFCT-77-101
September, 1977.
Fl J. S. Foster, E. Critoph, "Advanced Fuel Cycles in Heavy-Water
Reactors," AECL-5735.
Gl P. Greebler, General Electric Co., Private Communication, March,
1977.
G2 B. W. Gainey, "A Review of Tritium Behavior in HTGR Systems,"
GA-A13461, April, 1976.
HI L. C. Hebel, E. L. Christensen, F. A. Donath, W. E. Falconer, L.
J. Lidofsky, E. J. Moniz, T. H. Moss, R. L. Pigford, T. H. Pigford,
G. I. Rochlin, R. H. Silsbee, M. E. Wrenn, "Report to the American
Physical Society by the Study Group on Nuclear Fuel Cycles and
Radioactive Waste Management," Rev. Mod. Phys. 50, 1978.
12-2
-------
H2 S. R. Hatcher, S. Banerjee, A. D. Lane, H. Tamm, J. I. Veeder,
Trans. ANS 22, 334, 1974.
H3 D. R. Haffner, J. H. Chamber-Tin, T. M. Helm, D. R. Marr, R. W.
Hardie, and R. P. Omberg, "An Evaluation of Eight Sequential
Nuclear Non-Proliferation Options," HEDL, April, 1977.
H4 C. Hamilton, N. D. Holder, V. H. Pierce, and M. W. Robertson,
"HTGR Spent Fuel Composition and Fuel Element Block Flow,"
GA-A13886 Vol. I, Vol II - Appendix, July, 1976.
Jl M. S. Judd, R. A. Bradley, A. R. Olsen, "Characterization of
Effluents from a High-Temperature Gas-Cooled Reactor Fuel
Refabrication Plant," ORNL-TM-5059, December, 1975.
Kl P. R. Kasten, et al., "Assessment of the Thorium-Fuel Cycle in
Power Reactors7^ ORNL/TM-5565, January, 1977.
K2 C. W. Kee, A. G. Croff, and J. 0. Blomeke, "Updated Projections
of Radioactive Wastes to be Generated by the U. S. Nuclear Power
Industry," ORNL/TM-5427, Decenber, 1976.
LI R. K. Lane (General Atomic), Private Communication, July, 1976.
Ml D. A. Menelay, "CANDU Systems," ASME-ANS International Conference
on Advanced Nuclear Energy Systems, March, 1976.
M2 R. A. Matzie, J. R. Rec and A. N. Terney, "An Evaluation of
Denatured Thorium Fuel Cycles in Pressurized Water Reactors,"
ERDA Contract EY-76-02-2426, Presented at ANS Annual Meeting,
June, 1977.
M3 M. Maeda, Japan Atomic Energy Research Institute, Private Commun-
ication, May, 1977.
M4 F. M. Mann and R. E. Schenter, "Production of Uranium-232 in a
1200 Mw(e) Liquid-Metal Fast Breeder Reactor," Nucl. Sci. Eng. 27,
544, January, 1970.
M5 W. Marshall, "Nuclear Power and the Proliferation Issue," Nuclear
News. 34-38, April, 1978.
PI T. H. Pigford, R. T. Cantrell, K. P. Ang, B. J. Mann, "Fuel Cycle
for 1000 Mw High-Temperature Gas-Cooled Reactor," EEED 105
(Teknekron), EPA Contract 68-01-0561, 1975.
P2 T. H. Pigford and K. P. Ang, "The Plutonium Fuel Cycles," Health
Physics, 29_, 451, 1975.
12-3
-------
P3 T. H, Pigford, and J. C. Choi, "Economics of Fuel Cycle Options in a
Pressurized Water Reactor," Trans. AfiS 27, 463, 1977.
P4 D. T. Pence, "HTGR Reprocessing Wastes and Development Needs,"
General Atomic Report GA-A13919, April, 1976.
P5 H. C. Paxton, "Los Alamos Critical Mass Data," LA-3067-MS, 1975.
P6 T. H. Pigford, "Start-up of First-Generation Fast Breeders
with Plutonium of Enriched Uranium," UCB-NE 3240. March, 1977-
P7 T. H. Pigford, and J. S. Choi, "Effect of Fuel Cycle Alternatives
on Nuclear Waste Management," Proc. Symposium on Waste Management,
CONF-761020, October, 1976.
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Mann, November, 1977.
SI N. L. Shapiro, J. R. Rex, and R. A. Matzie, "Assessment of Thorium
Fuel Cycles in Pressurized Water Reactors," EPRI NP-359, February,
1977.
S2 B. R. Sehgal, J. A. Naser, C. Lin, W. B. Loewenstein, "Thorium-
Based Fuels in Fast Breeder Reactors," Nucl. Tech. 35, 635, October,
1977-
Tl C. E. Till and Y. I. Chang, "CANDU Physics and Fuel Cycle Analysis,"
ANL RSS-Tm-2, May, 1977.
T2 C. E. Till, et al., "A Survey of Considerations Involved in In-
troducing CANDU Reactors Into the U. S.," ANL RSS-TM-1, Feburary,
1977.
T3 R. F. Turner (General Atomic), Private Communication, April, 1976.
Ul U. S. Nuclear Regulatory Commission, "Nuclear Energy Center Site
Survey - 1975," NUREG-0001-ES, January, 1976.
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Safeguards (Main Report)," OTA No. 48, June, 1977.
U3 U. S. Federal Regulations 10 CFR 20, Appendix B, Table II.
U4 U. S. Nuclear Regulatory Commission, "Final Environmental Statement
on the Use of Recycle Plutonium in Mixed Oxide Fuel in Light Water
Cooled Reactors (GESMO)," NUREG-0002, 1976.
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Center," GEAR-11367, January, 1976.
12-4
-------
Yl S. Yamashita, "Variations in Neutronic Characteristics Accompany-
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Japan, 7_, 341, July, 1969.
12-5
-------
13. Nomenclature
A atomic weight of uranium isotope
E fuel exposure
J total interstage flow in ideal close-separation cascade
L capacity factor
M mass of isotope in fuel charged to reactor
N total number of atoms of a radionuclide
Q separative work
RCG radioactivity concentration guide for ingestion, i.e., maximum
permissible concentration in water
T preprocessing cooling time
c
Tp time elapsed between reprocessing and fabrication of
recycled uranium-thorium fuel
TR fuel residence time in reactor
T post-processing storage time for recovered thorium
x atomic fraction of light isotope
a ideal separation factor
$ fraction of recovered thorium to be recycled with bred uranium
n overall thermal efficiency
A radioactive decay constant
<)> separation potential (Chapter 7), neutron flux (Chapter 9)
ip 228Th activity in irradiated thorium relative to 228Th activity
in natural thorium
a microscopic cross section
02 232Th
08 228Th
22 232U
13-1
-------
Appendix A: Storage Time for Thorium Recovered From
HTGR Fuel Reprocessing
Thorium recovered from reprocessing irradiated thorium fuel may
have to be stored prior to recycle to allow time for decay of 24.1-day
23Hh and 1.91-yr 228Th, which are formed by the reactions discussed
in Chapter 2. After the fuel is discharged, and prior to reprocessing,
the 23^Th activity decreases with time. However, the activity of
228Th may increase if 228Th is not in secular equilibrium with 232U
at the time of fuel discharge. Although the total of the 228Th and
231*Th activities decreases with time, the activity from 228Th
daughters is the most troublesome when chemically purified thorium
is being refabricated. The highly energetic betas from both 228Th
and 23l*Th chains give rise to large skin doses upon surface contact
with separated thorium, but the highly energetic (2.6 MeV) gammas from
the 228jh decay chain can result in serious dose rates even with semi-
remote fabrication techniques. Here we focus upon 228Th, which
controls the requirements for post-process ing storage of recovered
thorium.
When the separated thorium is to be eventually recycled and
blended with low-activity uranium streams, such as make-up U, the
activity of 228Th in recovered thorium after a preprocessing cooling
time T and a post-processing storage time TS is given by
c
"A08Tc
"(A.I)
08 refers to properties of 228Th
22 refers to properties of Z;!ZU
N(TR) refers to the total quantity in the discharge fuel
THOH- can be.recycle^fo^fabrication^^low-activity uranium
of 228yn in natural
thorium,
to properties (A.2)
we obtain
T = -
5 A08
(ft.3)
A-l
-------
For the reference HTGR reactor of Chapter 6 with discharge con-
centrations of (AN)W(AN)n? = 4.05 x 103,(AN)nR/(AN)n? = 1.70 x 103,
T = 150 days, and ^ = 5, we obtain uo
L.
Ts = 16.5 yr
for thorium to be used when fabricating fuel with make-up 235U. In
the HTGR about two thirds of the thorium is used to fabricate fuel
containing make-up or recycled uranium containing no 232U, so about
two thirds of the separated thorium would be subjected to the storage
time estimated above.
For that portion of the separated thorium Which is to be eventually
recycled and blended with the recycled bred uranium, less time for
thorium storage is necessary. A reasonable criterion is that the
thorium be stored for a sufficient period such that its 228Th activity
is equal to the activity of 228Th in the recycled uranium at the time
of fabrication. Ignoring cross-over and process losses, the recycled
bred uranium contains all of the 232U which was present in the dis-
charge thorium. If this recovered uranium has been stored for a time
Tr prior to fuel fabrication, the activity of 228Th in the uranium
i!
U
08N08
0-
(A.4)
Applying the above criterion, we equate the 228Th activity in the bred
uranium to the activity of 228Th in the fraction B of the recovered
thorium that is eventually to be recycled for fabrication with the
bred urani urn, i.e.,
(A.5)
Th
where (A0gNoa) is the activity of 228Th in thorium after times
TC and T . Combining Eqs. (A.I), (A.4), and (A.5), we obtain
TS=A
08
22(TR)A22
1 - e
"A08TF
(A.6)
A-2
-------
For the 235U-Th-fueTed reference HTGR reactor, g = 0.36.
Assuming TC = 150 days ana Tp = 60 days, we obtain:
Ts = 3.1 yr
for the recovered thorium to be used when fabricating fuel with
bred 233U. As the pre-fabrication time TV of uranium storage
increases, less time is required for thorium storage. For the
parameters listed above if the recovered uranium is stored for
166 days before fabrication, the 228Th activity in the uranium
becomes equal to that of 36% of the separated thorium, so no
additional time for thorium storage would then be required to
meet the 228Th criterion of Eq. (A.5).
A-3
-------
Appendix B: Tables of Actinicies in CANDU Fuel Cycles
NOTE: IN ALL APPENDIX B TABLES, a REFERS TO ALPHA-ACTIVITY, AND
3 REFERS JO BETA-ACTIVITY.
TABLE B.I Actinldes in the Fuel Charged To
The Natural Uranium - Fueled
CANDU Reactor (1000 Mwe, no
Reprocessing!/)
Uranium!^ 235
238
Total
kg/yr
9. 075x1 O2
1. 267x1 O5
1.276xl05
Ci/yr
1.946
4.222xlO]
a = 4.417X101
weight %
0.715
99.285
100.00
a/ 7.5 Mw-day/kg of U, 30.5% thermal efficiency, 80%
capacity factor, equilibrium fuel cycle.
b/ U is not included.
B-l
-------
TABLE B.2 Actinides In The Fuel Discharged
From The Natural Uranium-Fueled
CANDU Reactor a/ (1000 Mwe, no
Reprocessing li/)
kg/yr Ci/yr
UraniumC/ 235 2. 233x1 02 4. 788x1 0'1
236 l.OOSxlO2 6.361
238 1.258xl05 4.192x101
Total 1.261xl05 a =4. 876x1 O1
Plutonium^/ 238 5.312X10"1 9.475xl03
239 3.201xl02 1.963xl04
240 1. 227x1 O2 2.778x10^
241 3. 086x1 O1 3.470x10^
242 8.74 3.408
Total 4.829xl02 a =5.689xl04
6 =3. 470x1 O6
a/ immediately after discharge
b/ 7.5 Mw-day/kg U, 30.5% thermal efficiency,
factor, equilibrium fuel cycle.
c/ 234Us 237|j and 239u are not included.
d/ 236Pu and 243pu are not included.
weight%
0.18
0.08
99.74
100.00
0.11
66.28
25.41
6.39
1.81
100.00
80% capac
B-2
-------
TABLE B.3 Actinides In The Fuel Charged To The
1.2% 235U-Fueled CANDU Reactor
(1000 Mwe, no Reprocessing!/)
kg/yr Ci/yr weight/E
Uraniumk/ 235 5.577x1O2 1.195 1.20
238 4.592x10^ 1.530x1Ql 98.80
Total 4.648x1O4 a =1.650x1O1 100.00
a/ 21 Mw-day/kg U, 30.5% thermal efficiency, 80% capacity
factor, near-equilibrium fuel cycle.
b/ 234u is not included.
B-3
-------
TABLE B.4
Uranium^/
Total
Plutonium^/
Total
235
236
238
238
239
240
241
242
Actinides In The Fuel Discharged From
The 1.2% 235U-Fueled CANDU Reactor!/
(1000 Mwe, No Reprocessing^/)
kq/yr
3.496X101
7. 563x1 01
4.507xl04
4.518xl04
1.748
1.209xl02
8.604X101
2.315x10"!
1.788x10'
Ci/yr
7.474xlO-2
4.796
1.502X101
a =1. 989X101
3.080xl04
7. 41 7x1 O3
1. 949x1 O4
2. 602x1 O6
6.978x101
weight%
0.08
0.17
99.75
100.00
0.70
48.42
34.45
9.27
7.16
2.498x1O2
a =5.778x1O4
B =2.602x106
100.00
a/ immediately after discharge
b/ 21 Mw-day/kg U. 30.5% thermal efficiency, 80% capacity
factor, near-equilibrium fuel cycle.
c/ 234U, 237u and239U are not included.
d_/ 236Pu and 243Pu are not included.
B-4
-------
TABLE B.5 Actinides In the Fuel Charged To The
U-Fueled CANDU with Self-Generated Pu
Recycle (1000 Mwe, with Pu Recycle §/)
Ci
Uranium^/
Total
235
238
Plutonium?./ 239
240
241
242
Total
kg/.yr
3.742xl02 8.023X10-1
5.234x1Q4 1.744xlQl
5.271xl04 a =1.824x101
1.706x1O2
1.610xl02
4.109x101
1.054x102
4.781xl02
1.046x104
3.647xl04
4.620x1O6
4.111xl02
a =4.734xl04
6 =4.620x1O6
weight%
0.715
99.285
100.00
35.68
33.68
8.59
22.05
100.00
a/ 18 Mw-day/kg U + Pu, 30.5% thermal efficiency, 80%
capacity factor, near-equilibrium fuel cycle.
b/ 234y is not included.
c/ 150 days cooling of discharge fuel before reprocessing
0.5% loss in reprocessing, 0.5% loss in fabrication.
and 238pu are not included.
B-5
-------
TABLE B.6 Actinides in the Fuel Discharged
From The U-Fueled CANDU with Self-
Generated Pu Recycle
-------
TABLE B.7
Actinides In The Fuel Charged To The
"bU-Th-Fueled CANDU Reactor
(1000 Mwe, with U recycle I/)
Thorium 232
Total
Uraniumk/
232
233
234
235
236
238
kg/yr Ci/yr
3.426x1O4 3.746
3.426xl04 a=3.746
110x101
674x1O2
673x1O2
_.879x1O2
1.923xl02
8.286X101
2.377xl03
4.430x1O3
1.035xl03
6.172x10-1
1.220x101
2.761xlO'2
Total 1.198xl03 <*=7.855xl03
weight^
100.00
100.00
0.01
39.02
13.97
24.03
16.05
6.92
100.00
a/ "27 Mw-day/kg Th+U, 30.5% thermal efficiency, 80%
capacity factor, near-equilibrium fuel cycle.
b/ 150 days cooling of discharged fuel before reprocessing.
0.5% loss in reprocessing, 0.5% loss in fabrication.
B-7
-------
TABLE B.8
Actinides In The Fuel Discharged
From the 235U-Th-Fueled CANDU
Reactori/ (1000 Mwe, with U RecycleE/)
Thorium£/ 232
Total
Protactinium*!/
233
kg/yr
3.335x1O4
3.335x1O4
3.393x101
Total
Uraniums./ 232
233
234
235
236
238
Total
Plutonium!/ 238
239
240
241
242
Total
3. 393x1 O1
1.41X10-1
4. 359x1 O2
1. 744x1 O2
7. 842x1 O1
2.1 02x1 O2
7.219X101
9.712xl02
6.643
2.176
8. 390x1 O'1
5.681x10-'
3.019X10-1
1. 053x1 O1
Ci/yr
3.647
a =3.647
7.042x108
3 =7.042x108
2.443x1O3
4.131xl03
1.079x1O3
1.681X10-1
1.333x1O1.
2.406x1O"2
a =7.667x1O3
1.161xl05
1.335x1O2
1.900x1O2
6.388xl04
1.177
a =1.164xl05
3 =6.388x1O4
weight%
100.00
100.00
100.00
100.00
0.01
44.88
17.96
8.08
21.64
7.43
100.00
63.10
20.67
7.97
5.39
2.87
100.00
a/ immediately after discharge
b/ 27 Mw-day/kg Th+U, 30.5% thermal efficiency, 80% capacity
factor, near-equilibrium fuel cycle.
c/ trace quantities of other Th isotopes are not included.
d/ trace quantities of other Pa isotopes are not included.
e_/ 237U and 239U are not included.
f/ 23&Pu and 243pu are not included.
B-8
-------
TABLE B.9
Actinides In The Fuel Charged To The
Pu-U-Th-Fueled CANDU Reactor
(1000 Mwe, with U Recycle!/)
kg/yr
Thorium 232 3.445x1O4
Total
Uranium!!/
Total
Plutonium^/ 239
240
241
242
Total
3.445x1O4
6.886x102
2.092xl02
8.019X101
2.014x101
5.713
3.152xl02
Ci/yr
3.767
=3.767
a =7.675xl03
1.283x1O4
1.816x1O4
2.265x106
2.228x1O1
a =3.101xl04
3 =2.265x106
weight^
100.00
100.00
232
233
234
235
236
1. 089x1 O'1
4. 606x1 O2
1. 577x1 O2
3. 990x1 O1
3. 033x1 O1
2. 332x1 O3
4. 365x1 O3
9. 760x1 O2
8.554xlO-2
1.924
0.02
66.89
22.90
5.79
4.40
100.00
66.36
25.44
6.39
1.81
100.00
a/ 27 Mw-day/kg Th+LH-Pu, 30.5% thermal efficiency, 80% capacity
factor, near-equilibrium fuel cycle.
b/ 150 days cooling of discharge fuel before reprocessing.
c/ 236pu and 238Pu are not included.
B-9
-------
TABLE B.10 Actinides in the Fuel Discharged
From The Pu-U-Th-Fueled CANDU
ReactorfL/ (1000 Mwe, with U Recycled/)
kg/yr
Ci/yr
weight%
Thorium^/ 232 3.356xl04 3.670
Total 3. 356x1 O4 a =3.670
Protactinium^/
233 3.345x10' 6.942xl08
Total 3. 345x1 O1 0 =6. 942x1 O8
Uranium§/ 232 1.119X10;1 2. 396x1 O3
233 4.293xlOz 4.068xl03
234 1.651x102 1. 022x1 O3
235 4.257X101 9.127X10'2
236 3.941X101 2.499
Total 6. 765x1 O2 a =7. 489x1 O3
Plutonium!/ 238 2.136 3.733xl04
239 1.637x10' 1. 004x1 O3
240 6.038x10' 1. 368x1 O4
241 1.91 9x1 O1 2.1 58x1 O6
242 1.815x101 7. 079x1 O1
Total 1.162xl02 a=5.208x!04
B=2.158xl06
§_/ immediately after discharge
b/ 27 Mw-day/kg Th+U+Pu, 30.5% thermal efficiency
capacity factor, near-equilibrium fuel cycle.
c/ Trace quantities of other Th isotopes are not
d/ Trace quantities of other Pu isotopes are not
e/ 23^U is not included.
f/ 236Pu and 243Pu are not included.
100.00
100.00
100.00
100.00
0.02
63.46
24.41
6.29
5.82
100.00
1.84
14.08
51.95
16.51
15.62
100.00
, 80%
included.
included.
B-10
-------
Appendix C: Calculational Methods
1. Light-Water Reactors
Cycle-by-cycle burnup calculations by Shapiro, et al. (SI) for
1330 Mwe PWR power plant fuel cycles operating on both uranium and thorium
fueling with segregated recycle were used to derive the material quantities
for the fuel cycles. The lattice code "CEPAK" was used in doing the point
(zero dimensional) reactor calculations. This computer code is a synthesis
of a number of other codes: "FORM", "THERMOS", and "CINDER", where "FORM"
is for the epithermal resonance and fast calculations on a homogenized cell,
"THERMOS" calculates the thermal spectrum for a one-dimensional representation
of the fuel cell, and "CINDER" does the fuel burnup calculations in a critical
spectrum calculated by "THERMOS" and "FORM", The spectrum calculations were
repeated prior to each burnup calculation to account for the spectrum effects
of the depletion of the fuel isotopes and the build-up of fuel and fission
product parasitic absorbers. The excess reactivity for leakage and control
margin was assumed to be 4%.
The material quantities were scaled according to the power level ano
were corrected to a capacity factor of 0.8. The lifetime-average quantise;;
shown in the mass flow sheets were calculated by accumulating the cycle-by-
cycle quantities over the reactor lifetime. The equilibrium cycles were
calculated from the data for the last reload designed for full burnup.
The computer code "ORIGEN" was used to calculate the radioactivity a;id
toxicity of the high-level wastes. The initial actinide quantities in the
high-level wastes were obtained from the discharge fuel concentrations from
the "CEPAK" outputs.
Lifetime ore requirements were calculated by accumulating the ore
requirements of each cycle over the reactor lifetime. For the first generation
fuel cycles, the initial core inventory was also included. For those cases
involving recycle of fissile material, and/or supply of fissile material
recovered from fuel reprocessing, the reactor was assumed to be run on the
slightly enriched (3%) uranium or on the 235U-Th fuel cycle until sufficient
fissile inventory was accumulated with the reactor and fuel cycle so that
the reactor could then operate on the equilibrium fuel cycle.
2. High-Temperature Gas-Cooled Reactors
Data for the reference HTGR flow sheet were adapted from the detailed
calculations of the concentrations of the nuclides in the various HTGR fuel
streams published by General Atomic Co. (H4). In their calculations, the
"GARGOYLE" code was used to calculate the flux spectrum and to perform the
burnup and activation calculations in nine energy groups (five fast and four
thermal). The core was represented as a point by using the core average
nuclide concentration, and'the "GARGOYLE11 code was used to determine the
core average neutron spectrum in each group with core leakage introduced as
positive or negative contributions to the fission source in each group. The
C-l
-------
nine-group cross sections were collapsed from the ENDF/B-IV file by the "MICROX"
code, which calculates the correct spectrum from the nuclide concentrations
and lattice geometry.
All the numerical values were based upon the ninth fuel reload, loaded
into the reactor at the beginning of the tenth year of operation, the last
reload presented in the GA report. These GA data were calculated on the
assumption of no cross-over between fissile and fertile streams in fuel
reprocessing.
For the fuel compositions of the subsequent reloads, the effective one
group cross sections deduced from the GA data were used. If we assume the
effective one group cross sections are constant from reload to reload, then
the discharge fuel concentrations of any nuclide can be expressed as linear
combinations of the initial nuclides concentrations, and constants can be
calculated from the ninth reload data and be used to calculate the discharge
fuel concentrations for the later reloads.
OOC
It was found that the concentration of U in the recycle bred uranium
fuel does not reach equilibrium even during the lifetime of a second generation
HTGR which has started up with the reactor and fuel cycle inventory of the
first generation HTGR. Because uranium equilibrium occurs so late after the
introduction of HTGR's, we chose to, concentrate on the fuel cycle deduced from
the General Atomic data from the ninth reload.
235
Because of a higher than normal amount of make-up U charged into the
reactor on the earlier (fourth) reload, which, after later discharge and
reprocessing, is fabricated to form the first-recycle 235y fue] Of ^ne ninth
reload, so there is a considerable perturbation in the fuel charged to the
ninth reload as compared with previous and subsequent reloads,. Therefore, the
data shown in the flow sheet were obtained by back extrapolation from the
later reloads to make the discharge concentrations vary monotonically from one
reload to another.
The effect of cross over on the fuel compositions was also calculated by
using the constant effective one group cross sections method described above.
3. Heavy-Water Reactors
The goals for the calculation were to first determine the critical fuel
composition at the beginning of cycle, and from this composition to determine
the end-of-cycle discharge-fuel composition. The computer code "EPRI-CELL"
(C7) was employed to do these calculations. "EPRI-CELL" is a computer code
very similar to the "CEPAK" used in the PWR calculation by CE. It also has
three built-in modules to calculate the space, energy and burnup dependent
neutron spectrum within a cylindrical cell. "GAM" solves the Boltzmann
equation to calculate the flux values for each of the 68 groups in the
epithermal and the fast range. Nuclides can be specified in the input to
receive heterogeneous resonance treatment; other nuclides in the cell will
be treated homogeneously. "THERMOS" computes the thermal neutron spectrum
(35 groups) as a function of position in a cell by solving numerically the
integral transport equation with isotropic scattering. After the "GAM" and
"THERMOS" calculations for one time step, the nuclide number densities, the
cross sections of those nuclides included in the cell calculations, and the
C-2
-------
neutron spectrum are all passed on to "CINDER" to perform the depletion
calculation. There are 20 depletion chains for 30 distinct heavy elements
and 69 decay chains for 179 distinct fission products in the "CINDER"
library. After each depletion calculation, the nuclide number densities
are returned to "GAM" and "THERMOS" to perform the spectrum calculation
for the next time step.
The heavy-water reactor cell used in the calculation is an equivalent
cell to the actual 37-element CANDU fuel bundle, where there were 8
alternating fuel and coolant concentric rings followed by the coolant/
calandria tubes and the moderator region.
The excess reactivity allowance for leakage, Xe override and
control margin was assumed to be 3.5% (Tl). Therefore, after each "EPRI-CELL"
calculation, the infinite multiplication factor is calculated and tested. If
it does not equal to 1.035, a new initial fuel composition is guessed and
the whole calculation is repeated. The figure below shows the flow diagram
of the calculation.
Go to the
next time step
s this the
last time step?
r
EPRI-CELL
I
C-3
-------
TECHNICAL REPORT DATA
friease read Instructions on the reverse before completing)
EPA 520/6-78-008
2.
3. RECIPIENT'S ACCESSION NO.
,. TITLE AND SUBTITLE
Thorium Fuel-Cycle Alternatives
5. REPORT DATE
November 1978
6. PERFORMING ORGANIZATION CODE
OITHORtS)
T. H. Pigford and C. S. Yang
8. PERFORMING ORGANIZATION REPORT NO.
UCB-NE-3227
9. PERFORMING ORGANIZATION NAME AND ADDRESS
Department of Nuclear Engineering
University of California
Berkeley, California 94720
10. PROGRAM ELEMENT NO.
11. CONTRACT/GRANT NO.
68-01-1962
12. SPONSORING AGENCY NAME AND ADDRESS
Office of Radiation Programs-Las Vegas Facility
U.S. Environmental Protection Agency
P. 0. Box 15027
Las Vegas, NV 89114
13. TYPE OF REPORT AND PERIOD COVERED
14. SPONSORING AGENCY CODE
EPA 520/6
15. SUPPLEMENTARY NOTES
16. ABSTRACT
Actinide material quantities and lifetime uranium ore requirements are
calculated for thorium fuel cycles in pressurized-water reactors, high-temperature
gas-cooled reactors, and pressure-tube heavy-water reactors, and are compared with
similar quantities for reference uranium-piutonium fueling in light-water reactors
and in fast breeders. Flowsheets are presented for national-international fuel
cycles for safeguard controls, including dispersed national reactors fueled with
thorium and denatured uranium. Long-term radioactivity properties of high-level
radioactive wastes are compared. Also compared are the production of 1L*C, 3H,
232U, and other activated radionuclides from these reactors and fuel cycles.
17.
KEY WORDS AND DOCUMENT ANALYSIS
DESCRIPTORS
b.IDENTIFIERS/OPEN ENDED TERMS
1ight-water/gas-cooled
reactors, nuclear fuel
processing, nuclear fuel
resources, nuclear power
economics, nuclear explo-
sive safeguards, nuclear
weapons proliferation,
nuclear fuel mqmt.strategy
c. COSATl Field/Group
1810
0702
1807
1001
0618,
1809
Nuclear fuel cycles
Thorium
Radioactive wastes
Nuclear electric power generation
Radiation hazards
Nuclear materials management
1312
18. DISTRIBUTION STATEMENT
Release unlimited
19. SECURITY CLA
Unclassified
168
20. SECURITY CLASS (Thispage)
Unclassified
22. PRICE
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