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Assessment of the Thorium Fuel Cycle
in Power Reactors

P.R. Kasten -

F.dJ. Homan
E. J. Allen K.J. Notz
D. E. Bartine' A. R. Olsen:
W. L. Carter R. H. Rainey
E. H. Gift J. E. Rushton
J. D. Jenkins M. L. Tobias
A. L. Lotts _

 

 

 

 

 

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OAK RIDGE NATIONAL LABORATORY

(A @\7 UNIONIEARBID EHGORRORATIONEEO RATHEENERGYRRESERR(EH DRVELORMENTRADMINLSYIRATION
 

Printed in the United States of America. Available from
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OAK RIDGE NATIONAL LABORATORY

OPERATED BY
UNION CARBIDE CORPORATION -
NUCLEAR DIVISION

UNION
CARBIDE

POST OFFICE BOX X
OAK RIDGE, TENNESSEE 37830

October 3, 1979

To: Recipients of Subject Report

 

Report No: ~ ORNL-5565 o Classification: Unclassified:
Author(s):: R. W. Swindeman |

 

Subject: Analysis of Creep-Rupture Data for Reference Heat
of Type 304 Stainless Steel (25-mm Plate).

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ORNL/TM-5565
Distribution
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Contract No. W—7405-eng—26
ASSESSMENT OF THE THORIUM FUEL CYCLE IN POWER REACTORS

P. R, Kasten
Central Management Offices

. F. J., Homan
A. L. Lotts
A, R, Olsen
Metals and Ceramics Division

E. J. Allen
*J. D. Jenkins
"J. E. Rushton
M. L. Tobias
Engineering Technology Division

.D. E. Bartine
Neutron Physics Division

.W. L. Carter

K. J. Notz

R. H. Rainey
Chemical Technology Division

- E. H. Gift
Oak Ridge Gaseous Diffusion Plant

Date Published: January 1977

OAK RIDGE NATIONAL LABORATORY
Oak Ridge, Tennessee 37830
operated by
UNION CARBIDE CORPORATION
for the

 

ENERGY RESEARCH AND DEVELOPMENT ADMINISTRATIOi

OB

 
 

 
Abstract - . - - - . . - . »®

Summary . . . . 4 0 4 s o4 .

Conclusions . . . .
Recommendations . . .

1. INTRODUCTION . .

2. PRESENTATION OF REPORT .

3. PERFORMANCE OF THORIUM AND URANIUM FUEL CYCLES IN

THERMAL REACTORS .

4. PERFORMANCE OF THE THORIUM AND URANIUM FUEL CYCLES

IN FAST REACTORS . .

.

CONTENTS

5. CONCLUSIONS AND RECOMMENDATIONS

5.1 Conclusions . . . .
5.2 Recommendations . .

Appendix A — Physics Considerations .

Appendix B — Thorium Fuel Cycles in LWRs

.

Appendix C — Thorium Fuel Cycles in HTGRs .

Appendix D — Thorium Fuel Cycle in HWRs (CANDU)
Appendix E — Thorium Fuel Cycle in FBRs .

Appendix F — Use of 233y and 238U in Fast Breeder

-

*

»

Reactors ("Denatured" Fuel Cycles) .

Appendix G — Ore and Separative Work Requirements in an
Integrated Nuclear Economy .

Appendix H — Reprocessing Cost Estimates

Appendix I — Fabrication and Refabrication Cost Estimates .

Appendix J — Institutional Considerations .

Appendix K — Studies and Programs Required to
"Americanize'" the CANDU System .

Appendix L — Summary of Calculations and Calculational

Methods .

iii

Page

vii

xii
Appendix M — Qualitative Overview of Recycle Process
Status for Various Reactor Systems .

Appendix N — Power Cost and Ore Utilization Summary .

Appendix 0 — Irradiation Performance of Thorium-
Containing Fuels . . « « « + « « « &

Appendix P — Comments on Fissile Availability for FBR
Economy . . ¢ ¢« o ¢ ¢ o o o o o « o o o

Appendix Q — Considerations Regarding Break-Even
Breeders . « ¢ ¢ « o o ¢ o 4 o o o s o e

iv

P-1

Q-1

 
Abstract

A study was conducted at Oak Ridge National Laboratory to evaluate
the role of thorium fuel cycles in power reactors. Three thermal reactor
systems were considered: Light Water Reactors (LWRs); High-Temperature
Gas-Cooled Reactors (HTGRs); and Heavy Water Reactors (HWRs) of the
Canadian Deuterium Uranium Reactor (CANDU) type; most of the effort was
on these systems. A summary comparing thorium and uranium fuel cycles in
Fast Breeder Reactors (FBRs) was also compiled.

Relative to thermal reactors, the results show that better UjOg
utilization is possible using thorium fuel cycles than can be achieved
with uranium cycles. However, thorium cycle use does not change the need
for FBRs so long as significant increases in nuclear power generation are
needed for long times., Commercialization of thorium cycles, including
recycle, would give added flexibility to the U.S. nuclear industry to deal
with any delay in FBR introduction or with commercialization of a low-gain
FBR. Further, thorium fuel cycles under certain circumstances can produce
lower cost power than uranium cycles, particularly at higher U30g prices.
Generally, it appears more desirable economically to recycle plutonium
with thorium rather than with uranium in thermal reactors. However, limi-
tations on the amount of available plutonium would significantly limit
overall improvements in fuel utilization. '

The most direct vehicle in which to take advantage of the improved
fuel utilization capabilities of the thorium fuel cycle are LWRs since
they will be generating most of the nuclear power during the next two
decades. However, the thorium cycle does not appear to compete economically
in present LWRs even at uranium prices over $100/1b. Of the thermal reactors
and under reference conditions of this study, thorium fueled HTGRs and HWRs
have the best fuel utilization performance, while HTGRs offer the best
opportunity for economic use of the thorium cycle. HWR(Th)s have about
the same fuel utilization characteristics as HTGRs, but at a higher power
cost.

In FBRs, thorium or thorium/uranium cycles provide a more negative
void coefficient of reactivity than does the uranium cycle; further,
mixed cycles provide an alternative fuel in the event that a full recycle
plutonium economy is limited. The use of metal fuel provides the best
nuclear performance from thorium cycles, and the superior physical
properties of thorium metal relative to uranium might lead to an economic
FBR with high fuel-utilization characteristics. The use of thorium in
FBRs can provide desirable fuel for both thermal and fast reactors while
increasing the ratio of thermal-to-fast reactors that can be maintained
in an FBR economy.
 

 

 
SUMMARY

A study is made of the role that thorium fuel cycles can have in
power reactors based on present estimates of economic factors, U30g
resources, and nuclear power growth scenarios. In doing this, fuel-
utilization characteristics and power costs are estimated for various
reference reactor types, treating both the uranium and the thorium fuel
cycles to obtain the relative merits of the different systems. Three
thermal reactor types are considered: Light;Water Reactors (LWRs),
High~Temperature Gas—-Cooled Reactors (HTGRs), and Heavy-Water Reactors
(HWRs). For these systems, benefits to be obtained by the introduction
of the thorium fuel cycle are evaluated on the basis of the relative
energy generation from a given U30g resource and on economic performance
as a function of U30g and uranium enrichment costs. Overall economic
benefits or penalties were estimated using a 7.5%/year discount factor.
A summary of the performance of thorium, uranium, and mixed fuel cycles
in Fast Breeder Reactors (FBRs) is also prepared, with both Liquid-Metal
Fast Breeder Reactors (LMFBRs) and Gas—-Cooled Fast Reactors (GCRFs) being
treated.

This study considers that there are no restrictions on either fuel
use or on fuel recycle and also determines the relative economic and
fuel-utilization performance of the thorium and uranium fuel cycles
in the various thermal reactor types. The evaluation criteria are
based primarily on economic performance, although U30g utilization
is also given importance. In determining economic performance, U30g
prices are varied from $25/1b to $300/1b. On the above bases, the use
of the thorium fuel cycle rather than the uranium cycle in thermal
reactors results in better U30g utilization and, in some cases, improved
economic performance. At the same time, if FBRs are introduced on planned
schedules, the use of LWRs and FBRs on the uranium cycle gives better
U30g utilization in a growing economy than does the use of the thorium
cycle in thermal reactors. However, if FBR introduction i1s delayed

gignificantly, the use of thorium fuel cycles is advantageous from a

fuel-utilization viewpoint.

vii
In the above context, the application of the thorium fuel cycle rather
than the uranium cycle is justified on the following bases. In thermal
reactors, the thorium fuel cycle permits: (1) more energy to be extracted
from U30g, thus providing a contingency position if commercial introduction
of the LMFBR is delayed; (2) more economic power generation than that from
LWRs (uranium cycle), particularly at higher U30g prices; (3) a decreased
burden on FBRs relative to early expansion needs when FBRs are first
introducted into the power economy; and (4) a higher ratio of thermal-
to-fast reactors in an established FBR power economy. In fast reactors,
the thorium or mixed thorium/uranium cycle permits: (1) a more negative
void coefficient in the core of the reactor, (2) the use of a "denatured”.
fuel (one in which uranium containing less than 20% fissile is the initial
fissile fuel) in selected reactors, and (3) production of a fuel which
has desirable features for both fast and thermal reactors.

The LWR provides the most direct route for application of the thorium
fuel cycle; however, the urani&m cycle in LWRs is more economic than the
thorium cycle for the reference conditions. Further, the estimated impact
of the LWR(Th) (with 1980 introduction) in improved fuel utilization is
less than that of either the HTGR or HWR (with 1995-2000 introduction);
also, use of the thorium cyéle in LWRs at an early date impacts the
production of Pu for early use in FBRs, while similar use of HTGRs or
HWRs at a later date does not.

Of the thermal reactors investigated, and for the reference evaluation
conditions, only the HTGR is more economic with the thorium cycle than with
the uranium cycle at present nuclear fuel costs. If the uncertainties
regarding commercial introduction of the HTGR in the U.S. can be resolved
favorably, then the HTGR appears to offer the best combination of economics
and fuel utilization performance with the thorium fuel cycle. While HTGRs
probably cannot be commercialized in time to put these advantages to wide
use before about 1995-2000, their impact on improving fuel utilization can
still be significant. The HWR is the next best system for thorium applica-
tion, having about the same fuel utilization characteristics as the HTGR
but higher power costs; again, this reactor type is less commercialized in

the U.S. than the LWR. In fast reactors, thorium or mixed fuel cycles in

viii

 
LMFBRs appears attractive for obtaining improved void coefficients of
reactivity, for use if metallic fuels are practical, and for use if
"denatured" fuel cycles are mandatory. |

The practical application of the thorium fuel cycle requires the
development of fuel recycle capability. In particular, the lack of
thorium fuel recycle capability has severe economic impacts on LWR(Th)s
and HWR(Th)s. While much technology already exists upon which future
work can be based with regard to fuel recycle development, considerable
effort is still needed relative to providing a practical demonstration
of recycle technology. Demonstration of recycle fuel irradiation
performance is also needed.

With regard to the application of plutonium fueling in thermal
reactors, Pu/Th appears economically attractive relative to Pu/238U;
further, Pu/Th appears economically preferable to 235y/Th fueling if
Pu costs are those associated with recovery from LWR fuel. At thé same
time, the concentration of fissile plutonium in fuel discharged from
natural-uranium HWRs appears to be too low to be economically recovered;
use of an enriched uranium cycle in HWRs changes that situation. Overall,
while Pu/Th fueling in thermal reactors appears economically desirable,
such fueling has only a small influence on improving fuel-utilization
performance, because of limited Pu availability. The primary justifica-
tion for Pu/Th use is an economic one and dependent on Pu price.

Specific unit costs are estimated for fuel fabrication, reprocessing,
refabrication, and associated operations; these are utilized with-
estimates of capital costs and operating and maintenance costs to give
power costs. For Uj30g prices less than approximately $40/1b, the lowest
power costs for thermal reactors are generally calculated when no fuel
recycle takes place, considering all fuel cycles. However, increasing
the U30g price makes fuel recycle the most economic option, and its
application increases the energy extraction from a given U30g resource.

For estimated reactor growth scenarios, thorium cycle use in LWRs
(CR ~ 0.7) provides 12 to 167% more energy, while HTGRs or HWRs (with a
CR v 0.8) provides about 20 to 50% more energy, based on thorium reactors
being introduced commercially on a large scale about 1995-2000. The
increase can be larger if a conversion ratio of 0.9 is employed, and

much larger if break-even breeders are utilized; however, in the HWRs

ix
and HTGRs examined, such high conversion ratios generally lead to
substantially increased fuel inventories as well as high fuel recycle
costs, so that the associated economic performance is unattractive.
Overall, for the reference conditions, economic benefits relative to
LWR(U)s with Pu recycle (and discounted at 7.5%/yr) are greatest for
HTGRs; based on HTGR capital costs being equal to LWR capital costs,

U30g prices of $100/1b, U30g resources of 2.5-3.5 million tons, and

an HTGR conversion ratio of 0.8, benefits are $6.4-21.6 billion.
Corresponding benefits are $1-3.8 billion for HWRs; a penalty is
associated with use of LWR(Th)s. Also, the discounted capital invest-~
ment in separations facilities appears significantly less for LWRs and
HTGRs than for HWRs. The above HTGR benefits do not take into considera-
tion the cost of developing commercial HTGRs. Estimating the cost of
developing HIGRs at $2 billion (undiscounted), an increase in HTGR
capital costs of $95-115/kW(e) cancels the benefits stated above. If
costs for developing HWR(Th)s are $0.5 billion (undiscounted), an increase
in HWR(Th) capital costs of $13-18 kW(e) cancels the HWR(Th) benefits

stated above.
Conclusions

1. Developing of the thorium fuel cycle is justified on the bases
of better U30g utilization, improved potential for long-term economics,
and additional flexibility with regard to fuel recycle alternatives.
Thus, introduction of the thorium fuel cycle provides additional power
generation capability in case of delayed introduction of commercial
FBRs, or in case there is introduction of a low-gain FBR on the reference
schedule.

2. Use of LWR(Th)s rather than LWR(U)s will increase the amount of
energy generated from a given U30g resource by about 20% above the reference
value, considering substitution of thoria for urania in present type LWR
designs. Use of LWR(Th)s beginning in 1995-2000 increases the energy
generation from specified U30g resources by 12*16% }élative to complete
use of LWR(U)s. However, LWR(Th) systems do not appeaf\économic compared

to LWR(U) systems based on present commercial reactor designs even when
\\

the U30g price is $100/1b or more.

 
3. If the uncertainties regarding commercial introduction of the
HTGR in the U.S. can be resolved favorably, then the HTGR appears to
offer the best combination of economics and fuel utilization with the
thorium fuel cycle. Further, possible future increases in_thermal
efficiency through application of combined cycle HTGRs significantly
increases economic and fuel utilization potential.

4. The HWR(Th) system appears better suited than the LWR(Th)
system for attaining high conversion ratios. However, the capital
component of the HWR power cost appears at least as high as that of
LWRs, exclusive of the HWR requirement for heavy water, such that total
power costs of HWRs appear higher than that of LWRs for U30g prices
less than v$50/1b. A decrease in HWR capital costs appears important
to HWR application in the U.S. At $100/1b U30g, the HWR(Th) system
is more economic than either the LWR(Th) or LWR(U) systems.

5. The use of HTGRs and HWRs with conversion ratios in the 0.8 to 0.9
range increases energy generation from a given U30g resource by 20 to 64%7,
considering the introduction of these reactors by 1995-2000. (Power growth
scenarios utilized in estimating the above considered nuclear power levels
to rise to 400 to 600 GW(e) by the year 2000.)

6. Operation of thermal reactors on Pu/Th fueling appears to be
economically attractive when Pu is recovered from LWRs or enriched-uranium
HWRs. However, the use of Pu/Th fueling does not have a large impact on
fuel-utilization characteristics because of limited Pu availability.
Further, the use of Pu in this manner does not permit Pu to be available
for startup of FBRs. The Pu needs of FBRs under reference introduction
and growth scenarios are such that reserving Pu for FBRs precludes
large-scale use of Pu/Th fuel cycles.

7. The economic application of the thorium cycle in thermal reactors
generally requires the establishment of a fuel recycle industry, particularly
for LWRs and HWRs (fuel recycle is also required for utilizing product
Pu and uranium from the uranium cycle). Without fuel recycle, the thorium
cycle can be used most effectively in HTGRs; however, recycle in HIGRs
is desirable to increase fuel-utilization performance, and is also
economically desirable when U30g costs rise above about $40/1b for the

reference conditions of this study.

xi
8. Converter reactor operation with conversion ratios above about
0.9 does not appear economical; the high fuel recycle costs associated
with low fuel burnups and the high fissile inventory requirements out-
weigh the improvement in fuel utilization achieved.

9. The discounted economic benefits from thorium cycle use in the
various reference-type reactors, and relative to LWR(U)s with Pu recycle,
vary from $1-3.8 billion for HWRs, and from $6.4-21.6 billion for HTGRs,
based on capital charge equality for LWRs, HTGRs and HwRS, economic conver-
tion ratios, estimated power growth scenarios, thorium reactor introduction
by 1995-2000, a U30g price of $100/1b, and U30g resources of 2.5-3.5 million
tons. The use of thorium cycles in reference LWRs does not appear economic
relative to use of LWR(U)s.

10. The HTGR economic benefits given in (9) above are cancelled if
the unit capital costs of HTGRs are increased by $95-115/kW(e) above
those for LWRs; similarly, the HWR economic benefits are cancelled if the
relative HWR capital costs are increased by $13-18/kW(e).

11. Although the nuclear performance of the thorium fuel cycle in
FBRs is generally not as good as the uranium cycle, use of mixed cycles
in FBRs may be satisfactory and/or desirable. Also, use of metallic fuels
might be possible with thorium while not with uranium because of the
superior properties of thorium-based metal relative to uranium alloys.
Use of metallic thorium fuel improves the performance of the thorium fuel
cycle relative to use of oxide fuel; however, safety considerations may
influence the use of metallic fuel.

12. From a safeguards viewpoint, developing a mixed uranium/thorium
fuel cycle permits a "denaturing" of recycle fissile fuels, since 433U
can be diluted with 238y, This flexibility could be important if fuel
feed to certain FBR power stations were limited to ~20% enriched uranium.

\\

Recommendations

 

Strong support should be given to the thorium~cycle HTGR as the best
contingency reactor in case there is a significant delay in the introduction
of LMFBRs. The emphasis should be placed on commercializing HTGRs and
introducing them on a time schedule such that they can capture a large

share of the nuclear power market by 1995-2000.

xii

 
In addition to the above, continued studies and evaluations should be
carried out on the LWR(Th) and HWR(Th) systems. Since LWRs are the most
direct vehicle for thorium utilization, LWR(Th) systems should be studied
more thoroughly to be sure the conclusions of this limited study are valid.
Such studies should also evaluate LWR designs based on a more advanced
technology; in the later case, care must be taken to consider the influence
of licensing requirements on design. Relative to heavy water reactors,
HWR(Th)s should be considered for commercial introduction into the U.S.
as a backup to the HTGR. Associated effort should include an evaluation
of the costs and schedules for commercial introduction and licensing of
HWRs, of the capital investments required for D,0 separation plants, and
an assessment of all costs not expected to be borne by industry.

Finally, it should be noted that the above considers no limitations
on fuel use or on fuel recycle other than those associated with economic/
technical factors. .If\limitations on fuel recycle are imposed upon the
nuclear industry because of safeguards considerations, the use of thorium/
uranium mixed cycles in FBRs may be necessary to have a breeder economy,
and could have implications on thermal reactor fuel cycles. This situation
was examined only peripherally in this report; based on the results obtained,
it is recommended that thorium/uranium fuel cycles in FBRs be studied in

detail, along with their possible interactions with thermal reactors.

xiii
1. INTRODUCTION

This report provides the results of a concentrated effort over a
two—-month period to assess the relative economic and fuel-utilization
performance of thorium and uranium fuel cycles in various reactor types.
Under the limited time condtions of this study, it was not possible to
go into the detail that would be desirable. Further, we were dependent

primarily on information which was quickly available to us, such as that

from specific organizations performing work on the concepts of interest

and from open literature publications and meeting presentations. With
regard to LWRs, relatively little detailed information regarding the
performance of the thorium fuel cycle was initially available; reactor
physics information utilized was largely obtained from the open literature,
along with results of independent calculations at ORNL; information was
also obtained from General Electric (ERDA-sponsored study) and from
Combustion Engineering (EPRI-sponsored study). In general, much of the
open literature information on thorium cycles in LWRs appears inconsistent
and optimistic with regard to use of the thorium cycle. The calculations
we performed were generally consistent with the Combustion Engineering
results and provided the basis for our evaluation of the thorium fuel
cycle in LWRs. At the“same time, the detailed analyses performed by GE
indicate that the economic use of Pu with thorium in LWRs is much more
complicated than presented by the relatively simple calculations utilized
in this brief study; such aspects were not considered in this report.

The HWR(Th) and CANDU results are largely based on studies performed
by Canada, as reported by Argonne National Laboratory. Because of their
interest, Canada has studied a wide number of uranium and thorium fuel
cycle cases; as a result, this particular reactor type was studied in
considerable detail relative to the comparative performance of thorium
and uranium fuel cycles.

For the HTGR, results obtained by General Atomic were utilized. ORNL
has been involved in HTGR development for many years and is quite familiar
with this concept and its performance; only those cases and parameter

values which were felt to be significant in evaluating the concept were
presented. No comparison is given between the thorium and uranium fuel
cycles for this reactor type, since with fuel recycle (the case of interest
in this study), the thorium cycle is preferred from both economic and fuel
utilization viewpoints.

Relative to FBRs, general information is based primarily upon open
literature publications; in addition, some detailed, specific calculations
relating to the relative performance of thorium and uranium fuel cycles
in FBRs (LMFBRs and GCFRs) were carried out, with emphasis on the use of
fissile uranium of less than 20% enrichment.

One of the important items influencing the calculated performance
of a given reactor type is the reactor physics analysis of the core.

It shofild be noted that not all of the reactor physics information used
here appears to be of the same quality. Specifically, we have confidence
that the results presented for the physics performance of the HTGR are
of high quality and are fair representations of what can be expected
experimentally. We have less confidence in the reactor physics data
given for the high-conversion-ratio HWR(Th) systems and believe that
the fissile fuel inventories tend to be low at the high conversion
ratios. With regard to the LWR(Th)s, our confidence in the results

was not great during the first part of this study, with the stated
performance considered to be opfiimistic. Since that time we have
modified our results, and more recently we have obtained detailed
information from Combustion Engineering on their comprehensive study

of thorium use in pressurized water reactors; that information largely
confirms the results given in the present report.

The reactor plant capital costs, operating and maintenance costs,
and associated economic bases used in this study are based on what we
believe to be consistent relative values for developed industries,
based on evaluations by others. The estimates of fuel fabrication,
refabrication, and reprocessing costs were based on consistenf
evaluations which we performed during this study, considering specific
flowsheets, processes, and equipment. The resulting economic factors

are termed the "reference" values employed in this study, and on which

 
the results are based. At the same time, based on past experience,
such reference cost estimates could have significant uncertainties.
The influence of uncertainties in relative costs on the results were
not treated in detail, but a few specific cases were treated.

This report initially contained an evaluation of the Light Water
Breeder Reactor (LWBR) and its prebreeders; however, the LWBR design
available to us did not represent the most recent design concept.

As a result, ERDA-DNRA requested that the LWBR not be included at

this time; they plan to evaluate the updated LWBR design at a future
time. Also, the report does not consider molten-salt reactors. Further,
relative to the reactors treated, only "reference-type" designs were
considered in obtaining the results since these designs have passed
through the licensing process. While designs other than reference can
be considered, a much more detailed study would be required, since
safety considerations would also have to be treated in detail. Specifi-
cally, LWRs could be redesigned to give improved nuclear performance at
the expense of safety margin or by depending upon advances in heat transfer
and fluid flow technology. Such designs were not considered, and no
specific conclusions are reached concerning their economic and fuel-
utilization performance.

An initial draft version of this report was widely distributed during
July and August of 1976 to obtain review comments concerning this study.
The responses obtained have been most helpful and were carefully considered
during the preparation of the present report. As a result, a number of
small technical changes were made; additional discussions have been provided
to make the intent of certain sections more clear, and there have been
changes in the overall presentation to make the report more readable.

We believe this study to be a significant initial effort with
meaningful overall results on which future work can be based; at the
same time, there are considerable uncertainties in important economic

and parameter estimates and in certain '"ground rules." For example,

this study generally assumed that there are no restrictions on fuel use
or on fuel recycle other than those dictated by technical and economic
considerations. Changing that ground rule would drastically alter the
interaction of fuel cycles. Thus, there is a continuing need to factor
in new information as it develops relative to the evaluation of the

role of thorium fuel cycles in power reactors.

 
2. PRESENTATION OF REPORT

Because of the diverse nature of this study and the limited time
period during which most of the information was developed, compiled, and
written, various topics were treated separately, and these are presented
in the appendices. The report proper makes use of the general results
obtained in the separate studies, although additional evaluations were
also performed. 1In general, the material in the appendices gives more
detailed information concerning the specific subject matter. At the
same time, because the evaluation approaches utilized in the wvarious
appendices are not always the same, differences in detail exist between
the report proper and certain appendices. Sections 3 and 4 below
provide the final evaluations.

The attached appendices provide an information resource concerning
various aspects of this study; a brief description of their content is
given below. |

Appendix A summarizes the reactor physics aspects of thorium fuels
in both thermal and fast neutron spectra. The purpose of this appendix
is to provide perspective relative to the reactor physics features
associated with the use of various fuels in power reactors, since these
features influence the desirability and practicality of thorium fuel
cycles.

Appendices B, C, and D describe detailed studies performed on
thorium fuel cycles in LWRs, HTGRs, and HWRs. The results in Appendix B
are largely based on open literature publications in conjuction with
additional information obtained from various sources; both uranium and
thorium fuel cycles were treated. It was difficult to get a consistent
evaluation of the thorium fuel cycle based on the open literature publi-
cations; Appendix B compares the various results and places them in
perspective. Appendix C summarizes thorium fuel cycles in HIGRs and
the variation in performance associated with various core design features.
Results were largely obtained from General Atomic. Appendix C also treats
the operation of HTGRs as near-breeders, considering prismatic and pebble-

bed type fuel elements. Appendix D summarizes the performance of thorium
fuel cycles in HWRs of the CANDU reactor type, based largely on Canadian
designs., The information in Appendix D determined that the optimum
conversion ratio for the plutonium/thorium fueled concept is about 0.9,
while the optimum conversion ratio for the 235U/Th fueled concept is
about 0.8.

Appendices E and F concern fast breeder reactors; Appendix E
summarizes open literature information on the thorium and uranium fuel
cycles in FBRs. Appendix F provides results on specific calculations
performed relative to the use of 233U/238y in FBRs to examine the feasi-
bility of using uranium containing less than 20% fissile material in
LMFBRs or GCFRs.

Appendix G illustrates ore and separative-work requirements in an
integrated nuclear economy based on estimated reactor parameters and
specified nuclear power growths,

Estimates of the cost of fuel recycle are given in Appendices H and I.
These estimates are based on specific flowsheets, on equipment require-
ments, and on operating requirements, with special emphasis on consistency.
The unit costs for reprocessing various reactor fuels are given in Appendix
H, along with unit costs for fuel shipping and waste storage. Similarly,
Appendix I provides unit cost estimates for fuel fabrication and refabri-
cation. Appendix M gives a qualitative overview of the fuel recycle
process technology required for the various reactor systems, and also
estimates the sequential fuel recycle development costs for the various
reactor types.

Appendix J gives a brief discussion of some of the institutional
considerations associated with the introduction of the thorium fuel
cycle into the nuclear economy; Appendix K summarizes some of the studies
and programs required to "Americanize' the CANDU systems. This latter
appendix considers only R&D requirements and does not include demonstra-
tion programs or those programs that might be required as a result of
licensing studies.

Appendix N summarizes the power costs and ore-utilization performance
of the various thermal reactors, based on information given in Appendices
B, C, and D, and utilizing the calculational methods summarized in Appendix
L. Not all the economic parameters used in this appendix are the same as

given in Sections 3 and 4 below.

 
Appendices 0, P, and Q give useful additional information in areas
related to this study. Appendix O summarizes the irradiation performance
of thorium-containing fuels for the various reactor types. Appendix P
considers fissile availability in an FBR economy based on a specific
power growth rate in which LWRs are utilized initially. The influence of
HTGR use is also considered. Appendix Q gives an analysis of the HTGR
while operating as a near- or break-even breeder.

The relative performance of the thorium and uranium cyclés in thermal
and fast reactors are evaluated in Sections 3 and 4 below. For the
thermal reactors, the relative fuel-utilization and economic performance
of the two fuel cycles are considered for LWRs, HWRs, and HTGRs, considering
several power growth scenarios and Uz0g resources. In evaluating thermal
reactors, it is implicit that only the thermal reactor component of a
nuclear power industry be considered. So long as fissile material
(assumed to be Pu) is stored for later use in FBRs, it will always be
possible to initiate an FBR industry which can grow with time. For
the fast reactors, nuclear performance is summarized relative to the
use of thorium or uranium fuel cycles; in addition, some consideration
is given to the use of denatured fuel cycles. Section 5 gives the

conclusions and recommendations based on Sections 3 and 4.
3. PERFORMANCE OF THORIUM AND URANIUM FUEL CYCLES
IN THERMAL REACTORS

In evaluating the role of thorium fuel cycles, important criteria
are fuel-utilization and economic performance (including associated
economic benefits or penalties). In this section, the various thermal
reactors considered are assessed with regard to their relative energy
extraction from a given ore resource under various nuclear power growth
scenarios, along with associated power costs, and economic benefits or
penalties relative to a reference power cost. These evaluations also
treat the influence of uranium-ore and separative—-work prices on
power costs as a function of reactor type and fuel cycle operation.
Thermal reactors of the LWR, HWR, and HTGR type and of reference design
are considered here.

Two basic nuclear power growth scenarios are considered; in one,
power capacity growth is maintained at 15 GW(e)/year up to a level of
450 GW. After the power level reaches 450 GW, it is maintained at that
level until it is necessary to reduce the capacity because of limitations
in U30g resources. In the second scenario, nuclear power growth occurs
at 30 GW/year until the capacity reaches a level of 900 GW(e). The
power capacity is maintained at 900 GW(e) until it is necessary to
reduce the level because of limitations in U30g resources,

The power growth scenarios are indicated in Figures 1, 2, 3, and 4.
In all cases, reference-type LWRs, termed LWR;'s, are utilized initially;
after a given time, new reactors are built. The new reactors are either
LWR,'s (same as LWR; but identified differently to clarify results),
LWR(Th)s (LWRs operating on the thorium cycle), HWRs, or HTGRs. For
Scenario I given in Fig. 1, LWR;'s are installed at a rate of 15 GW(e)/
year from 1970 until the year 2000; after that time, they are withdrawn
from use as their 30-year lifetime is attained. The LWR;'s withdrawn
from use are replaced with a second type reactor as indicated above.

As shown in Fig. 1, the power capacity is maintained at 450 GW(e)

for a period of time, to, defined as the time of extension associated
with maintaining the power capacity at 450 GW(e). After time t,, no new
reactors are built, and those in use are operated until the end of their

30-year lifetime.

 
ORNL-DWG 76-19369

 

 

 

 

 

 

 

 

 

 

 

 

 

500
' |
~ e to= TIME OF
——————tm——\ EXTENSION
\
400 \
@ \
3, \
LWR2 or
= 300 2 \-
§ HTGR or \\
S
HWR or \
® T 200
W R(Th
- LWR, LWR (Th) \
S \
I Z \
100 ‘.\
\
\
o \
1970 2000 2030 2060 2090
YEAR

Fig. 1.

Thermal reactor power growth scenario I (Initial growth

of 15 GW(e)/year; LWR,, LWR(Th), HWR, or HTGR introduced in 2000).
10

ORNL-DWG 76-19370

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

500
- fe
r———-“
/ \

400 ¢ \
T A
3. \
z 300 LWR, or \
Q
< LWR (Thyor \
S \
© HTGR or \
ax 200 Y
< HWR \
-
3 \
2 LWR, \

100 \

\
0 \
1970 2000 2030 2060
YEAR
Fig. 2. Power growth scenario IA (Initial growth of 15 GW(e)/year;

LWR,, LWR(Th), HWR, or HTGR introduced in 1995).

 
11

ORNL-DWG 76-19374

 

 

 

 

 

 

1000
--—————fe-————l-1
900 [~~~ \[te=TIME OF
/ \ EXTENSION
800 ,' X
f \
700 £ \
\
/
600

 

 

LWRj or l\
\ LWR (Th) or \

\ HTGR or  \
\ HWR —

/ LWR, \ \\
200 \
/ \l
100 \T
[ \
0 \

1980 2000 2030 2060
YEAR

Fig. 3. Reactor power growth scenario II (Initial power growth of
30 GW(e)/year; LWR,, LWR(Th), HWR, or HTGR introduced in 2000).

 

3 8
\

 

 

 

 

NUCLEAR CAPACITY [GW(e)]
9
3

 

 

 

 

 

 

 
NUCLEAR CAPACITY [GW(e)]

12

ORNL-DWG 76-19372

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1000 I |
ot te
50 /'"'-"""""\\ te= TIME OF
800 il . EXTENSION
/ \
700 / \
/ \
/ \
600 \
/ LWR2 or \
500 1 \
/ \ LWR(Th)or| \
\
400 / \ HTGR “or \
300 HWR “L
/ LWR, \ \
200 / \ \
\
100 \ \
/ \ \
0 \

 

1980 2000 2020 2040 2060 2080
YEAR

Fig. 4. Power growth scenario ITIA (Initial power growth of
30 GW(e)/year; LWRy, LWR(Th), HWR, or HTGR introduced in 1995).

 
13

A variation in the above power growth scenario, termed Scenario
IA, is also considered and shown in Fig. 2. 1In Scenario IA, LWR;'s
are installed from 1970 to 1995 at 15 GW(e)/year; no new LWR; reactors
are built after 1995, and those in operation continue until the end of
their 30-year life. Starting in 1995, a second reactor type is built
(either LWR,, LWR(Th), HWR, or HTGR) such that the power capacity
increases to 450 GW in the year 2000; the power capacity is then
maintained at that level until it is necessary to shut the reactors
down because of limitations in U30g resources. Again, the time during
which new reactors are introduced, including the time the power capacity
remains at 450 GW(e), is called t, (time of extension); t, is indicated
in Fig. 2.

The second type of power growth scenario, termed Scenario II, is
shown in Fig. 3. 'In this specific case, the power capacity increases
from zero in 1980, at a rate of 30 GW(e)/year, until a level of 600
GW(e) is reached in the year 2000. The newly constructed LWR; reactors
are operated for 30 years, resulting in the power capacity curve shown
for LWR;. Starting in the year 2000, a second reactor is installed at
a rate of 30 GW(e)/year, such that the total power level rises to 900
GW(e) by the year 2010. After that time, the power level is maintained
at 900 GW(e) until the U30g commitment associated with a 30-year reactor
life equalé the U30g resource. The time during which new reactors are
installed, along with the time at which the power capacity is maintained
at 900 GW(e), is again termed tes the time of extension, and is indicated
in Fig. 3.

A variation in the above power growth scenario, termed Scenario ITA,
is given in Fig. 4. In Scenario IIA, the time of growth of LWR;'s takes
place from 1980 to 1995, after which time new reactors are installed.

As shown, the maximum power capacity rises to 900 GW(e). Again, t,
designates the time during which new reactors are built and includes
the time during which the power capacity remains at 900 GW(e).

The above power growth scenarios, along with the U30g requirements
of the various reactors, permits the calculation of the energy that can
be generated from a given U30g resource. In calculating the mined Uj30g

needs, the reactor lifetime requirements given in Table 1 are employed.
14

Table 1. Relative U30g Needs?

[One 1000-MW(e) Reactor; 21 Full-Power Years,
0.2% Tails]

 

 

U30g%

Reactor Fuel Recycle (Tons)

LWR No 6010
LWR U only 4780
LWR U/Pu 3890
LWR(Th) Yes 3242
CANDU | No 4650
CANDU Yes 2710
HTGR (CR = 0.66) No 4310
HTGR (CR = 0.66) Yes 2680
HWR or HTGR (CR = 0.82) Yes 2032
HWR or HTGR (CR = 0.85) Yes 1925
HWR or HTGR (CR = 0.90) Yes 1774
HWR or HTGR (CR = 1.0) Yes 1710

 

aU308 requirements include initial core inventory; however, for HWR or
HTGR with CR = 1, the entire fuel cycle inventory is included.

The values in Table 1 are estimated U30g needs associated with one
1000-MW(e) reactor operating for 21 full-power years; also, the concen-
tration of 23°U in the "tails" from the enrichment process is taken to
be 0.2% 235U. The U30g requirements include the initial core inventory.
Although at first it will take more than the initial core inventory
because of the inventory associated with fuel fabrication and reprocessing,
all the inventory becomes available later as the plants are shut down.
However, the availability of that inventory may not always correspond
to U30g needs; for reactors with a fuel conversion ratio (CR) less than
unity, the core inventory is included to compensate for the possibility
that fuel-cycle-inventory availability may not be properly phased with

the U30g needs of the remaining plants as nuclear capacity is decreasing.

 
15

However, in the case of a break-even breeder, energy can be generated
indefinitely. As a result, it is necessary to include the inventory
needs of the entire fuel cycle, as is done in Table 1.

Based on the above nuclear power growth scenarios and the lifetime
U30g requirements associated with the various reactors, estimates of the
relative energy generated by the various reactors are obtained, as well

as estimates for t the years during which new reactors are being

e
constructed and the power level is being maintained. In obtaining those
estimates, it is always assumed that 1200 MT of the fissile Pu generated
by LWR; will be stored and saved for FBR use [this would be the quantity
of fissile PU reduired by FBRs in a power growth.economy of 30 GW(e)/year,
where the FBRs have a specific inventory of 4 kg fissile/MW(e) and an
overall fuel doubling time of 20 years (simple doubling time, including
the fuel cycle inventory]. Further, the ore resource considered available
for reactor use is either 2.5 million tons U30g or 3.5 million tons U30g.
The results obtained are given in Figs. 5, 6, and 7 for the various
power growth scenarios and reactor types. As stated previously, LWR,
represents a continued construction of LWR;, but is specifically identified
to clarify the results. As shown in Fig. 5, for power growth Scenario I
and a U30g resource of 2.5 million tons, LWRy's have a t, of 8.6 years.
Further, the energy generated by LWR; plus LWR; is considered as the
reference energy generation based on the use of LWRs (on the uranium fuel
cyle) to consume the entire ore resource (except for 1200 tons of fissile
Pu which is held in storage for FBRs). On that basis, the relative energy
generated by LWR; plus LWR, is unity, as indicated in-Fig. 5. Also shown
in Fig. 5 are the results when 3.5 million tons U30g are considered, as
given by the dashed lines. In this latter case, t, for LWR; is 25 years,
and again the relative energy generation for LWR; plus LWR; is unity. The
results for the various reactor systems are also shown, with "LWR(Th)"
representing LWRs operating on the thorium fuel cycle with recycle of bred

fuel.* "HTGR(0.66)" represents a steam—cycle HTGR operating with a fuel

 

X
Relative to LWR(Th) use, all the capacity of LWRs is converted to
LWR(Th)s at the time LWR(Th)s are introduced.
1.7

1.6

1.5

14

1.3

LWR, + LWRz

1.2

1.4

RELATIVE ENERGY GENERATION
(LWR, + LWR/HTGR/HWR)

1.0

0.9

20
18
16
14

12

tg, YEARS (for solid lines)

10

16

ORNL-DWG 76-19373

 

 

 

— 25 x10% tons U30g

 

 

== 35 x 0% tons UzOg

 

 

 

 

 

L 2 X L _J
—-_

 

 

 

 

 

 

 

o
s NS NS GED GEnb I SEND GRED GUN| SIIEP GEYP RS SIS GE
=¥ ¥ B K I N X X ¥ XK 1 _J

 

 

 

e |— S S
—x X X 2 X __J

 

 

 

 

 

 

 

 

 

8 8 8§ 8 & 3o

 

 

 

- il S EENS SN P

L1

 

A Sy GEES s EED GEND SED S
-t

e cEnn GElEy Sanb GEED SN S N

s cnn alee c— - —" D SEm SE—

e e e Gl oIS GG SE— G I
s s e aEn GED SIS SED S SEE A G S

 

 

 

 

 

 

 

 

 

0

 

8
LWR2

LWR(Th) HTGR  HTGR(066) HTGR/ HTGR(082) HTGR/ HTGR(0S0)
(066) CC AFTER HWR CCAFTER HWR CC AFTER
{0 YEARS (082) 40YEARS (090) 40 YEARS

REACTOR SYSTEM

Fig. 5. Relative energy generation for power growth scenario I
(15 GW(e)/year, new reactors in 2000).

te » YEARS (for dashed lines)

 
17

ORNL-DWG 76-19374

 

 

 

2.5 x 108 tons U30g

 

 

 

 

 

 

 

 

 

 

AUITITITIIIINUNININN

RELATIVE ENERGY GENERATION
(LWRy + LWR/HTGR/HWR)

 

 

 

ARUNITUIIIINNNINN

 

 

 

o
NUUINNNNN

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

09

34

30
7
" 26 / /
-
3 2 Z
7 7 7

14
LA O A 7

LWRZ LWR(Th) HTGR/HWR HTGR/HWR

(0.82) (0.90)
REACTOR SYSTEM

Fig. 6. Relative energy generation for power growth scenario TA
(15 GW(e)/year, new reactors in 1995).
18

ORNL-DWG 75-19375

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

16
< 0
8 1.5 !
%‘g . —— NEW REACTORS IN 2000 . |L
: T
gzl ——= NEW REACTORS IN 1995 | |
O e P
> g £ 43 |
|t
w Elg 42 + 1
Z + % 1 |
w == | i
w3 1 4 +
B | I
d 10 i— | iL
o
0.9 1 1 1 .
i
|
20 } 26
' o
—~ 18 J 1 26 8
3 I | £
= 16 ! | » B
2 | : 8
% 1 : + 0
&’ 12 % 18 g
g I <
w i I o
> ! :
8 — } 14
' } } |L

 

 

 

 

 

 

 

 

 

 

6 12
LWR3 LWR(Th) HTGR HTGR(066) HTGR/ HTGR(082) HTGR/ HTGR(0S0)

(0.66) CC AFTER HWR CCAFTER HWR CC AFTER
10 YEARS (0.82) 1{OYEARS (0.90) 10 YEARS

REACTOR SYSTEM

Fig. 7. Relative energy generation for power growth scenarios IIL
and IIA (30 GW(e)/year; 3.5 x 10° tonms U30g).

 
19

conversion ratio of 0.66 and with recycle of bred fuel; "HTGR(0.66)-CC
after 10 years'" represents the use of steam cycle HTGRs initially, with
combined cycle HIGRs (which employ gas turbine topping cycle and an
ammonia turbine bottoming cycle) utilized for all new reactor construction
10 years after the introduction of HTGRs; "HTGR/HWR(0.82)'" represents use
of steam-cycle HTGRs or HWRs operating on the thorium cycle, with either
reactor type having a fuel conversion ratio of 0.82:; "HTGR(0.82)-CC after
10 years" represents HTGRs with a conversion ratio of 0.82 and combined-
cycle HTGRs being built 10 years after HTGR introduction; similarly,
HTGRs and HWRs with a CR of 0.90 are also considered. The use of
combined-cycle HTGRs permits the additional generation of energy because
of the higher thermal efficiency of that system (efficiency is estimated
to be 487 compared with 397 for the steam cycle HTGR).

The results in Fig. 5 show that the relative energy generation is
significantly influenced both by reactor type and by the amount of U30g
available for use. Using 2.5 million tons of U30g, about 127 more energy
is obtained with LWR(Th)s rather than LWRs after the year 2000;* if HTGRs
or HWRs with a conversion ratio of 0.82 are employed, the relative energy
generation is 207 more than the reference value. 1If HTGRs with a conversion
ratio of 0.9 are utilized, with combined-cycle HTGRs employed 10 years .
after the introduction of steam—cycle HTGRs, the relative energy generation
is about 1307%. Alternatively, if the U30g resource is 3.5 million tons,
the latter value increases to 165% of the reference value. The times of
extension for the various cases are given in Fig. 5 and range from 8.6 to
66 years.

If new reactors are started in 1995 rather than in 2000, different
values for t, and relative energy generation are obtained. Figure 6
gives results for Power Growth Scenario IA (Fig. 2), with 2.5 million
tons of U30g being utilized; as shown, the relative energy generation is
1147% for the LWR(Th) case, rising to 1327 for HWRs or HTGRs with a CR
of 0.82 and to 141% for HWRs or HTGRs with a CR of 0.9. Similarly, the

 

*
Relative to LWR(Th) use, all the capacity of LWRs is converted to
LWR(Th)s at the time LWR(Th)s are introduced.
20

time of extension, te, varies from 19.5 to 30 years relative to a value
of 13.6 for the reference LWR. Comparing the results from Fig. 6 and
Fig. 5 indicates that introducing HWRs or HTGRs with a CR of 0.82 in
1995 instead of 2000 increases the relative energy generation from a
value of 1207% to 132%.

Figure 7 gives relative energy generation results for Power Growth
Scenarios II and ITA (initial growth of 30 GW(e)/year; new reactor types
in either 1995 or 2000) for a U30g resource of 3.5 million tons. With
new reactors introducted in the year 2000, the relative energy generation
by new reactors gives energy increases of 13% for LWR(Th)s, 257 for either
HTGRs or HWRs having a CR of 0.82, and 46% for HTGRs having a CR of 0.9,
along with the introduction of combined-cycle HIGRs 10 years after the
introduction of steam-cycle HTGRs. The value of te increases from 7.6
years for LWRs to 18.6 years for steam-cycle HTGRs. When new reactors
are introduced in 1995 instead of 2000, the relative energy generation
by new reactors gave energy increases of 16% (instead of 13%) for LWR(Th)s,
42% (instead of 25%) for HTGR/HWRs (0.82), and 55% (instead of 33%) for .
HTGR/HWRs (0.9). The values of t, increase from 12.7 years for LWRs to
28 years for HTIGR/HWRs (0.9).

In addition to fuel-utilization aspects, it is important to evaluate r
the economic performance of the various reactor types, This evaluation
is accomplished by treating uranium-ore and separative-work prices as
parameters for the various reactor types. The calculations performed
and cost factors employed are similar to those utilized in Appendix N
and are discussed below. The results given present a consistent evaluation
of the relative power costs in the various thermal reactor concepts as a
function of U30g costs, separative-work costs, and for consistent estimates
of the fuel fabrication, refabrication, and reprocessing costs. The unit
recycle costs take into consideration the throughput of the recycle plant
associated with a specific reactor concept, with some consideration given
to the influence of scale on unit costs as the throughput of the plant
is increased. At the same time, the costs of fuel recycle do not include
estimates for fuel shipping, storage, and waste treatment, as given in
Appendix H. The slightly lower effective recycle costs utilized here .

(relative to those given in Appendices H and I) give somewhat preferential

 
21

treatment to those reactors having relatively low fuel exposures. The
fuel cycle cost factors employed are listed in Table 2, which also gives
the U304 and separative-work prices that are considered, in addition to
the estimated cost of thoria. Specific values for the fabrication of
fresh fuel, for reprocessing of fuel, and for the refabrication of recycle
fuel are given for the various reactor concepts; these values represent
1976 cost estimates. Effective fuel storage costs used here are also
listed. The cost to recover fissile plutonium considers only that cost
associated with reprocessing the material. On the above bases, and
assuming that the first reactor cycle always pays for fuel storage, the
cost of recovering fissile Pu is about $20/g for LWRs and about $24 to
$50/g for HWRs. The above effectively assumes that the first reactor
fuel cycle will "write off" any fuel value of the product Pu.

Other general features of the fuel cycle cost calculations employed
in this section are given below. With the CANDU reactor (operating with
natural uranium), the fuel is .considered to be obtained for fuel fabrica-
tion 1/2 year before reactor exposure; in all other reactor concepts,
fuel fabrication is considered to require having fuel "on hand" one year
before reactor exposure. The time for fuel reprocessing and conversion
is considered to require fuel to be ''on hand" for one year following
reactor exposure, for all reactor concepts. In the case where fuel is
stored, the fuel and fabrication "inventory" is written off over the
reactor lifetime and an appropriate "inventory factor' is utilized to
properly account for those costs over the period of fuel exposure. When
fuel is recycled, two basic situations are considered; in one, the
first cycle is considered specifically; the second treats all subsequent
operations to be on the "equilibrium cycle." For the first cycle, the
fuel and fabrication "inventory' is considered to be written off over
the lifetime of the fuel; as a result, fuel which is recycled to subse-
quent cycles has no cost to those cycles other than costs associated
with fuel refabrication and reprocessing. As a result, for the equilibrium
cycles, the "inventory'" charge is only associated with the “makeup" fuel,
and with the write-off of working capital associated with fuel fabrication/

refabrication. The average fuel cycle cost is then obtained by averaging
22

Table 2. Fuel Cycle Cost Factors

 

A.

Ore and Separative Work Factors

U30g, $/1b/SWU, $/kg

 

25/75
40/100
100/150
300/200
B.
Reactors Fuel Fab.
ILWR
U235/U238 114
Pu/U
U235/Th 152
U233/Th
Pu/Th
HWR
Natural U 50
Enriched U 80
Pu/U
U%35/Th 100
U233/Th
Pu/Th
HTGR
U235 400
U233/Th
Pu/Th

Reactor Fuel

HWR (natural U)
HWR (enriched)
LWR

HTGR

LWR

ThO2
$30/kg

(no recycle considered)

Fuel Cycle Cost Parameters, $/kg

 

Fuel Reprocess.

221
221
250
250
260

150
160
160
210
210
220

750
750
730
750

C. Effective Fuel Storage Costs

Fuel Refabrication

 

500

570
510

310

390
320

1030

1030

Storage Costs, $/kg

25
100
100
400

D. Cost to Recover Fissile Pu

$20/g

$24=-550/g

 

 

 
.

-.

23

the fuel costs of the first cycle and the equilibrium cycles on the basis
that the reactor lifetime is 30 years and that a discount factor of 7.5%/
year applies. Other than the items mentioned above, the general calculation
of the fuel cycle costs employed the same methods as described in Appendices
N and L.

Table 3 lists the additional power cost factors which are employed
here to obtain power costs. As shown, the capital charge rate is 16%/
year (however, the capital charge rate relative to fuel cycle working
capital is taken to be 15%/year). The heavy water cost is taken to be
$110/kg, and the heavy water inventory of an HWR is considered to be
0.8 kg/kW(e); heavy water losses from HWRs are taken to be 2%/year.
Reactor operating and maintenance costs are taken to be a nominal 2
mills/kWhr(e), which is estimated to be the appropriate value for
approximately 1980. The capital cost of LWRs is considered to be the
reference basis for capital costs. An LWR unit capital cost of
$800/kW(e) is utilized here and is based on estimates for a plant
starting operation in the early 1980s.* The absolute value of the
capital cost is not so important in this study as the relative capital
costs for the different reactor types. Reasonable chénges in the above
LWR cost estimate would not have a significant influence on the_results
of this study so long as relative costs are correct. Thus, the use of
capital costs based on reactor operation in 1982-83 and of consistent
fuel recycle costs based on construction of recycle plants in 1976 still

permits a valid evaluation of thorium and uranium fuel cycles in the

different reactors.

The capital cost for én HWR uranium system considers that the unit
capital cost of an HWR operating at 80% load factor is the same as that
of an LWR operating at 75% load factor. These relative values are in
reasonable agreeement with the information presented by Argonne National
Laboratory in their draft 1976 report on HWRs, and also are consistent

with the relative cost information developed for LWRs and HWRs as reported

 

* .

W. K. Davis, "Economics of Nuclear Power," Proceedings of the Inter-
national Symposium on Nuclear Power Technology and Economics, Vol. I,
pp. 29-69, Taipei, Republic of China, January 13-20, 1975.
24

Table 3. Power Cost Factors

 

Capital Charge Rate: 16%/year (15%/year for fuel cycle)
D,0 Cost: $110/kg

D,0 Inventory: <~ 0.8 kg/kW(e)

D,0 Losses: 2%/year

 

 

 

 

Reactor Capital Cost Load 0&M D»0 Cost
Type $/kW(e) mills/kWhr (e) factor,? mills/kWhr(e) mills/kWhr(e)
LWR 800 19.5 75 2.0
HWR(U) 853 19.5 80 2.0 _ 2.26
HWR(Th) 843 19.3 80 2.0 2.0
HTGR-SC 800 19.5 75 2.0
HTGR-CC 720 17.6 75 2.0
in WASH-1087.% Further, the capital cost of the thorium-fueled HWR .

relative to the uranium-fueled HWR is somewhat lower due to the slightly

tighter lattice spacing that could be used for the thorium cycle system.

This difference is also reflected in the heavy water costs for HWR(Th) 7
systems, with the D,0 inventory and makeup costs reduced by about 107

relative to those costs for uranium-fueled HWRs.

The relative capital costs of the HTGR are based on recent evaluations
by United Engineers and Constructors,+ who estimated that the unit capital
costs for SC-HTGRs (when developed to the same extent as LWRs) were
essentially the same as those for LWRs. The CC-HIGR costs are taken to
be 107% less than those of the SC-HTGR costs (the UE&C cost estimate for
this system was about 15% less than the SC-HTGR).

 

*
Advanced Converter Task Force, 4An Evaluation of Advanced Converter
Reactors, WASH-1087, April 1969.

TUnited Engineers and Constructors, Inc., Gas-Cooled Reactor Assessment
for the Energy Research and Development Administration, Vol. II, "Capital
and Operating Costs — Safety and Environmental Assessments," June 22, 1976.

 

 
q .

25

The resulting calculated power costs for the various reactor systems
are summarized in Figs. 8-11. The calculated costs for the LWR afe
given in Fig. 8; as shown, results are given for the LWR operating on the
uranium cycle with storage of fuel and for the LWR with recycle of uranium
and plutonium; the latter case is considéred to give the reference power
cost against which new systems need to compete. Results are also given
for the LWR(Th), initially fueled with thorium and 235y and with recycle
of bred 233y,

With regard to the horizontal lines associated with Pu/Th or
Pu/2380U fueled systems, the term "limited" implies there is a limited
amount of Pu which is available., Further, the horizontal lines imply
that the power cost is independent of U30g cost. The Pu cost is that
cost associated with recovery from the first LWR uranium cycle, with
the '""fuel value" of the Pu being "written off" over the first cycle.

It can be noted in Fig. 8 that the use of Pu with thorium has a power
cost about 1.5 mills/kWhr(e) lower than the use of Pu with uranium based
on the estimates and calculations employed here.

Recycle Pu can have a value higher than the cost of recovery,
particularly if it is recycled soon after discharge from the reactor.
However, if the spent fuel is stored after exposure without certainty
of recycle, its value should be written off over the fuel exposure.
Further, since the uranium cycle in LWRs appears to be able to "write
off'" exposed fuel economically, acceptance of this procedure encourages
fuel recycle and reactor operation at relatively high fuel conversion
ratios.

As shown in Fig. 8, use of uranium cycle LWRs with storage of fuel
appears to be the most economic option for LWRs up to a Ujz0g cost of
about $50/1b. Above the cost, it appears more economical to recycle
uranium and Pu. However, use of LWR(Th)s with recycle of fuel does
not appear preferable to the uranium cycle even at high U30g costs.
Maintaining the cost of separative work constant at $100/kg SWU for
U30g costs above $40/1b would help the thorium cycle in LWRs. Us30g
costs/SWU costs of $100 per 1b of U30g/$100 per kg SWU instead of
$100/$150, respectively, decreases LWR(Th) costs by about 0.3 mills/

kWhr(e) relative to LWR(U) systems; similarly, employing costs of
42

40

38

W W
H N

POWER COST [milis/kWhr(e)]
O
N

30

28

26

26

ORNL-DWG 76-19376

 

 

 

 

LWR, U CYCLE STORE

 

 

 

 

 

 

 

1/ RECYCLE

 

 

| LWR-Th/2%%y / LWR, 2% + Py
RECYCLE ) / (Iimited)

  

 

 

 

- LWR,Th+ Py
. imire
/ fi. b

 

 

 

 

 

 

25/75  40/100 1007450 300/200
UsOg COST (8/1b)/SW COST (¥kg SWU)

Fig. 8. LWR power costs,

 
42

40

38

W W W
N H o0

POWER COST [mills/kWhr (e)]

W
O

28

26

24

27

ORNL-DWG 76-9847

 

 

 

 

 

/
Th/235y, CR=0.66, RECYCLE /
Th/235y, CR=0.82, RECYCLE\TZ/
/
/,
/

 

 

 

235 n — h
Th/¢35U,"CR=0.66, STORE “/fi /
/
/
/

 

1
/
/
Th/235y, CR=0.90, RECYCLE //
y

 

/ /'/
/A
/s Pu/Th, CR=0.90,

 

 

//// RECYCLE (LIMITED)

 

 

 

 

 

 

 

 

 

 

 

 

Fig. 9.

/4
=452/ Pu/Th, cR=0.82,
— RECYCLE (LIMITED)
— -
/
Pu/Th, CR=0.66,
RECYCLE (LIMITED)
|
25/75 40/100 100/150 300/200
U30g COST ($/1b)/SW COST ($/kg SWU)

HTGR power costs.
POWER COST [mitls/kwhr(e}]

44

42

40

38

36

32

30

28

26

24

28

ORNL—-DWG 76-9849

 

 

NATURAL U-STORE/}

{

/

 

/
/

 

/
/
/

 

 

/

I 235U/Th —
RECYCLE,

l CR=0.82

 

i
’

/

| /

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

/
NATURAL U+ Pu RECYCLE /
| (LIMITED) /
/

ENRICHED U/Pu )/ | _Pu/Tn, RECYCLE,
RECYCLE = ~ CR—0.9

/ (LIMITED

/‘
—
l__,.-—"
25/75 40/100 100/150 300/200

U30g COST (35 /1b)/SW COST (§/kg SWU)

Fig. 10.

HWR power costs.

 
29

ORNL-DWG 76-19377

 

 

 

 

 

 

 

      

 

 

 

 

 

 

 

 

 

 

 

40
38
/l
36 2
/P
34 / l.l
] /
2 /
=
2 p
.7 %
2 - 7/
= LWR-U/Pu RECY?LE\7,/
[ S— ] ,l
= LWR(Th)\\’/'
3 30 !
o
i
=
O
o
28
" /
— / SC-HTGR
7 CR=0.66
. A
- -
" CC-HTGR
CR=082
25/75 40/100 100/150 300/200

U3Og COST (¥/1b)/SW COST ( ¥kg SWU)

Fig. 11. Comparative power costs for CC-HTGR(GT) and HTGR, HWR,
and LWR.
30

$300 per 1b/$100 per kg instead of $300/$200 decreases the relative
LWR(Th) costs by about 0.5 mills/kWhr(e). Such changes would make the
LWR(Th) cycle more attractive at the higher U30g costs, based on the
evaluations given here. [Recent results by Combustion Engineering,
however, indicate that the above economic performance for LWR(Th)s
relative to LWR(U)s is optimistic.]*

Figure 9 gives estimated power costs for HIGRs as a function of
U30g and separative work costs. Results are generally for the thorium
cycle with recycle of the bred 233U; however, storage of spent fuel is
treated for the low CR design. Costs are also given for Pu/Th fueling
with recycle of bred 233y for various conversion-ratio designs; in these
cases there are a limited number of reactors which can be built because
of the limited amount of Pu which is available. The associated costs
are shown independent of U30g cost on the same bases given above for
Pu/Th use in LWRs. Because of its limited application, little emphasis
is given to the use of Pu, other than pointing out that Pu use with
thorium appears economically attractive. Overall, Fig. 9 shows that
up to a U30g price of about $40/1b, it is about as economic to store
fuel as it is to recycle fuel in the most economic reactor, which has
a conversion ratio of about 0.66. (In the case of fuel storage, the
CR is less than 0.66; however, in order to identify the specific core
design, the term "CR = 0.66" is used.) As the cost of U30g rises,
it becomes important to recycle fuel, and at $100/1b for U30g the cost
of power from an HTGR with a CR of about 0.82 is about the same as that
from a reactor with a CR of 0.66, based on the estimates and calculations
used here. At a CR of 0.9, however, the power cost does not appear as
favorable as with a CR of 0.82. At the same time, if low-cost Pu is
available, high CR systems appear economically attractive.

Figure 10 gives the estimated HWR power costs. The results indicate

that the natural uranium system with fuel storage is the most economic

 

* :
Private communication from Norton Shapiro, Combustion Engineering,

to Paul Kasten, ORNL, October 19, 1976.

 
31

one up to U30g prices of about $100/1b U30g. At U305 prices above about
$130/1b, the thorium cycle with a CR of 0.82 becomes more economic than
the natural-uranium cycle with fuel storage. However, use of natural
uranium plus Pu, with fuel recycle, is more economic than either of the
above cycles at the higher ore prices. (In this case, the number of
reactors operating on this cycle are limited by Pu availability.) The
use of Pu with Th appears economically attractive at U30g prices above
about $60/1b in reactors with a CR about 0.9. (Again, the cost is shown
to be independent of U30g price because Pu is considered to be available
from the first uranium cycle for only the cost of recovery; the number
of reactors which can be operated on this cycle is limited because of
limited availability of Pu.)

Pu can also be recycled in HWR uranium cycles; however, recycle
of Pu in natural uranium systems is not as economic as the natural-
uranium cycle with fuel storage, because of the refabrication penalty
associated with adding Pu to all of the fuel. If Pu is to be recycled
in uranium systems, it should be employed in conjuction with the enriched
uranium cycle, with the Pu utilized in only a fraction of the fuel
elements in order to reduce the effective fuel refabrication'penalty.

As indicated in Fig. 10, use of the latter cycle (enriched U/Pu, recycle)
gives power costs about the same as the thorium cycle, with thé thorium
cycle tending.to be lower at UzOg prices above about $40/1b.

Figure 11 gives a summary of selected cost information taken from the
previous figures so as to place power costs of the various reactor systems
in perspective, and in addition shows the economic performance of the
combined cycle HTGR (CC—HTGR)* with a CR of 0.82. It can be noted that
the CC-HTIGR system has significantly lower power costs than the other
systems; at the same time, it will take longer to introduce the CC-HTGR
commercially than the SC-HTGR.

Figure 12 summarizes power costs of thorium cycles in the different

reactors (with fuel recyle) relative to LWRs recycling uranium and Pu.

 

*
The CC-HTGR refers to an HTGR employing a gas turbine topping cycle
and an ammonia turbine bottoming cycle, with an overall thermal
efficiency of 48% (the steam cycle HTGR has an efficiency of 39%).
32

ORNL-DWG 76-19378

 

 

 

 

 

2
! .---’;"""s..__. (LWR
0 s m——
\. '\<‘HWR
-1 C \\ ~C

 

(CR=066)| & (CR=082)

 

 

 

 

 

CC-HTGR
(CR=0.82)

 

COST OIFFERENTIAL [mills/kWhr (e)]
5
AV
/
/

 

 

 

\

-11

 

 

 

 

 

 

 

25/75  40/100  100/450 300/200
U30g COST (¥/1b)/SW COST (¥/kg SWU)

Fig. 12. Power costs relative to LWR(U/Pu) (Recycle in all cases).

 
33

On that basis, the power cost differentials between a given thorium
system and the LWR with U/Pu recycle is given as a function of U30g

and separative work costs. The cost differentials given for the differnt
systems show that an HWR(Th) system with a CR of 0.82 does not become
economical relative to LWRs until U30g costs exceed about $70/1b, and

that the economic advantage at $100/1b U30g is only about 0.5 mills/kWhr(e).
The LWR(Th) system at a CR of approximately 0.7 does not become economical
at any of the U30g prices considered under the evaluation conditions
utilized. HTGRs are the most economical systems shown in Fig. 12, having
a power cost advantage relative to LWRs for all U30g prices employed; at
U30g prices of approximately $100/1b, the economic advantage of the steam-
cycle HTGR with CRs of 0.66 to 0.82 is about 3 mills/kWhr(e), whereas that
of CC-HTGRs is 6 mills/kWhr(e).

Based on the above cost bases, economic factors, and the associated
power cost differentials given in Fig. 12, the economic benefits of the
various systems can be calculated. In all cases, the benefits are
calculated relative to a reference power cost equal to the LWR (U/Pu
recycle) system. Figures 13 and 14 give the results of these calculations
using a discount factor of 7.5%/year tb obtain discounted benefits (back
to 1976), with Fig. 13 considering Power Growth Scenarios I and IA, and
with Fig. 14 considering Power Growth Scenarios II and ITIA. The term
"delta" in the above figures refers to the unit power cost of savings
associated with the specific system, and is relative to the cost of power
from the LWR (U/Pu recycle) system. It is assumed that the reference
cost of power always applies, even though the LWR may not always be
available; thus, if the reference power source alternative to the LWR
were to cost more than that of LWR (U/Pu recycle), the benefits obtained
would be higher than those shown. In calculating the discounted benefits
of future systems, it is assumed that the price of Uz0g will be $100/1b

at that time, and the relative cost differentials associated with that

price are used in calculating the discounted benefits shown in Figs. 13

and 14. The terms used to describe the various reactor systems are those

used previously.
ORNL-DWG 76-19379

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

REACTOR
SYSTEM
LWR, fi
LWR(Th) -
HTGR(0.66) A L
C T ]
HTGR(066)+ (et et e — = e =S e
CC AFTER
{O YEARS R s Y X L ey r A—=|-3 A.C_ie--.
1 12.2
HWR (082) TA4=05
X X X/ — - low
HTGR (0.82) ___________________,_T_
HTGR (0.82) + e Gy IS GED Ghe CENS GlES GEE S SIS GHue SIS S S - ap
CC AFTER T T
| f
= UNIT POWER COST SAVING mms/kwnr(e)]
= UNIT POWER COST SAVING OF CC-HTGR
I l | | l l [mitis/kwhr (e)]
| | |
—-— 25x10° tons Uz0g; NEW REACTORS IN 1995
—— 25x10° tons U30g; NEW REACTORS IN 2000
—— 3.?)(1061 tons U3Og; NEW REACTORS IN 2000
O 4 2 3 4 5 6 7 8 909
DISCOUNTED BENEFITS (dollars)
Fig. 13. Discounted benefits for power growth scenarios I and IA

(Initial power growth = 15 GW(e)/year).

 
35

REACTOR ORNL-DWG 76-19380
SYSTEM

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

LWR, !
LWR(Th) =
HTGR (066) =3
HTGR (066) + A=5c=3
CC AFTER
10 YEARS 8306
HTGR(082) e ———— ——-4=3 |1 _|26
HTGR(0.82) + B=8c=3
CC AFTER
A = UNIT POWER COST SAVING [mitls/kWhr (e)
Ac= UNIT POWER COST SAVING OF CC-HTGR
l ' l . . [mills/kwhr (e)]
| I
NEW REACTORS IN 2000
——— NEW REACTORS IN 1995
L 1 [ 1 [ [ ]

 

O 2 4 6 8 10 12 14 16 18 (x109
DISCOUNTED BENEFITS (dollars)

Fig. 14. Discounted benefits for power growth scenarios II and IIA
(Initial power growth = 30 GW(e)/year; U30g resource = 3.5 x 10% tons).
36

Figure 13 gives the benefits (discounted to 1976) for Power Growth

Scenarios I and IA (see Figs. 1 and 2), where the initial power growth

rate is 15 GW(e)/year, and new reactors are introduced in either 1995

or 2000.

Since LWR, in this case is LWR (U/Pu recycle), and provides

the reference cost, there are no benefits shown for LWR,. Similarly,

since the LWR(Th) system has higher costs than LWR(U) systems, there

are no benefits shown for LWR(Th)s. [At the same time, if the cost

of separative work is considered to be $100/kg SWU instead of $150/kg,
the LWR(Th) benefits are about $1.9 to $2.3 billion for Power Growth

Scenario

I and for Uj30g resources of 2.5 to 3.5 x 10% tons, based on

this study. In obtaining this benefit, all LWRs which are in service

in the year 2000 are converted to LWR(Th)s at that time; subsequently,

only LWR(Th)s are utilized.]

Similarly, results are given for the other reactor systems. By far

the most benefits are obtained with HTGRs, with

followed by CC-HTGRs having the most benefits. For HTGRs (CR = 0.82)

introduced in 2000, and with a U30g resource of 2.5 x 10° toms, the

benefits

10 years

are $6.4 billion. If a combined-cycle HTGR is introduced

after introduction of the steam cycle HTGR, and a unit power

cost savings of 3 mills/kWhr(e) is applied to all HTGRs, the discounted

benefits

are $6.7 billion. If, on the other hand, the CC-HTIGR had a

unit power cost saving of 6 mills/kWhr (e) rather than 3, the discounted

benefits
resource
increase

For

2000 and

are about $8.2 billion. For this latter case, if the U3zOg

is increased to 3.5 million tons, then the discounted benefits
to $12.8 billion.”

the HWR with a CR of 0.82, and with new reactors introduced in

a U30g resource of 2.5 x 10% tons, the discounted benefits are

estimated to be about $1 billion; increasing the U30g resource to

3.5 x 10® tons increases the benefits to $1.5 billion. Introducing

such reactors in 1995 (with a U30g resource of 2.5 x 10% tons) results

in benefits of $1.8 billion.

 

*
If the cost of separative work is $100/kg instead of $150/kg, the
above relative benefits will increase by about 8% for the SC-HTGR.

introduction of SC-HTGRs

 
37

Figure 14 gives results similar to those in Fig. 13, except Power
Growth Scenarios II and ITA are treated (see Figs. 3 and 4), where the
initial power growth rate is 3Q GW(e)/year, and new reactors are introduced
in either 1995 or 2000. In all cases a U30g resource of 3.5 x 10°® tons
is assumed. The results show no benefits for LWR(Th)s under the reference
evaluation conditions. Similarly, HTGRs with a CR of 0.82 show benefits
of $11.4 billion for introduction in 2000 and $21.6 billion for intro-
duction in 1995; analogous benefits for HWRs (CR = 0.82) are $1.9 billion
and $3.6 billion, respectively. Figure 14 also shows the benefits
associated with HTGRs having a CR of 0.66, and with introduction of CC-HTGRs.
Overall, HTGRs again show the largest benefits for the evaluated conditions.

The results in Figs. 13 and 14 show that it is important to bring in
a new system as early as possible in order to increase the benefits td be
obtained from that system (for the scenarios studied, relative benefits
increase by about 90% when new reactors are introduced in 1995 rather
than in 2000). On the other hand, the results also indicate that the
economic benefits from HTGRs introduced in 2000 are significantly greater
than the benefits from HWRs or LWR(Th)s introduced in 1995.

Another economic factor to consider is the capital investment for
separation facilities required with the various reactor systems. In
particular,.heavy water separation plants are required for HWRs, while
enriched-uranium reactor systems require uranium enrichment facilities.
Comparing the separation facility investments for natural-uranium HWRs
with those of LWRs and HTGRs indicates there is a higher discounted
capital investment required for HWRs than for either LWRs or HTGRs.
Specifically, based on estimated relative investments of $3 billion for
a uranium enrichment plant producing 107 kg SWU/year, and of $1 billion
for a heavy water plant producing 1000 MT D,0/year, and a nuclear power
growth rate of 30 GW(e)/year, the discounted capital investment associated
with the separations facility (employing a discount factor of 7.5%/year)
is about $12 billion for HTGRs, about $14 billion for LWRs (uranium cycle),
and about $24 billion for HWRs (natural uranium). Thus, the HWR requires
about $12 billion more in discounted capital investments for separations
facilities than does the HIGR. If the nuclear capacity growth were
15 GW(e)/year, the capital investments would be one-half of those above,

in which case the HWR would require $6 billion more than the HTGR.
38

Another factor to consider is the development costs associated with
introducing a new reactor concept. Developing a commercial HIGR could
cost more than developing a commercial HWR(Th) or LWR(Th). Specifically,
the anticipated development costs for SC-HIGRs are estimated to be about
$1.4 to $1.5 billion™ (including about $900 million for fuel recycle
Qevelopment and demonstration), with development of CC-HTGRs costing
about $500 million® more when developed with the SC-HTGR. Thus, the
total development costs of SC- and CC-HTGRs are estimated to be about
$2 billion. The discounted value of the above $2 billion would be about
$1.4 billion (at a discount factor of 7.5%/year). Thus, even though there
were no development costs associated with HWR(Th) or LWR(Th) systems, the
net economic benefits to be obtained with HTGRs for the reference conditions
are greater than from either of the other systems. Of course, there are
development costs associated with commercializing HWR(Th)s and LWR(Th)s;
it is estimated that costs of fuel recycle development plus those of a
demonstration fuel recycle facility would be about $600 million or more.
An uncertainty in HWR costs also involves in part the uncertainty in
design and development required to license HWRs in the U.S.

As shown previously in Figs. 11 and 12, the CC-HTGR has significantly
lower power costs than the other systems. This illustrates the economic
importance of increasing the thermal efficiency of a given reactor system
when doing so does not cause a corresponding increase in plant capital
costs, and does not require a new type, more expensive fuel system. At
the same time, the benefits for the CC-HTGR relative to the SC-HTGR are
dependent upon the discount factor employed and the time of introduction
of the reactor systems; the benefits shown in Figs. 13 and 14 are based
on introduction of CC-HTGRs 10 years after SC-HTGRs, and a discount
factor of 7.5%/year. As a result, the discounted benefits from CC-HTGR
use relative to SC-HTGR use are not as large as might be expected from
the results given in Fig. 11. Nonetheless, the benefits are still

significant, and justify estimated expenditures for CC-HTGR development.

 

*
A. D. Little, Inc., Gas-Cooled Reactor Assessment, Vol. II1, prepared
for ERDA, August 1976 (NTIS, Springfield, Va.).

 
39

The correct value to use for the discount factor is difficult to
determine. In commercial ventures, discount factors greater than 7.5%/
year are prevalent. However, for something so basic as the ability to
produce energy for long periods of time, values much less than 7.5%/year
can be appropriate. While we believe the value used is appropriate for
the evaluation performed, it should be recognized that there is significant
uncertainty as to the correct value to employ. A value lower than 7.5%/
year would give more weight to future benefits than given here, while a
higher value would give more weight to near-term benefits.

Although not treated here,'increasing the core specific power and/or
the mean energy of neutrons causing fission tends to help the relative
performance of the thorium cycle in LWRs. The basic questions related
to the development of such LWRs concern the permissible safety margins
and the associated heat transfer/fluid flow performance. Licensing

requirements are thus a key concern for such LWR designs.
40

4. PERFORMANCE OF THE THORIUM AND URANIUM FUEL CYCLES
IN FAST REACTORS

This section provides a summary of the performance of thorium,
uranium, and mixed fuel cycles in fast breeder reactors (FBRs), with
both Liquid-Metal Fast Breeder Reactors (LMFBRs) and Gas-Cooled Fast
Reactors (GCFRs) being treated. In these reactors, the use of thorium
or thorium/uranium fuel cycles provides a more negative void coefficient
of reactivity in the core than does the use of the uranium fuel cycle;
since the coolant void coefficient of reactivity is much larger in
LMFBRs than in GCFRs, the above effect is more important in LMFBRs.
However, the use of the thorium cycle in conjunction with ceramic fuels

leads to lower breeding ratios than does the use of the uranium cycle.

While oxide fuels based on the thorium cycle have slightly better material

and thermal performance properties than similar fuels for the uranium
cycle, such differences do not appear significant.

With regard to metallic fuels, the material and irradiation
performance properties of thorium-based metal alloys are more suitable
to reactor use than are uranium metal alloys. Thus, the use of metallic
fuels might be possible with thorium but not with uranium. Further, the
use of metallic fuels based on the thorium cycle leads to breeding ratios
comparable to those obtained with ceramic fuels on the uranium cycle.
From a thermal hydraulic viewpoint under steady-state conditions, thorium
metal fuels appear able to operate at higher heat ratings than do oxide
fuels; however, safety considerations may limit power densities in metal-
fueled systems. Also, while irradiation experience to date with thorium
metal fuels in encouraging, it is limited, and much more development work

is required before utilization of such fuels can become a reality.

It is evident that the recycle of fuels is required for fast reactors

to operate effectively as breeders. The development of fuel recycle
capability involves similar effort and demonstration for either the

uranium or thorium fuel cycle. The present effort is on the uranium

cycle; the inclusion of thorium cycles will require an incremental increase

in effort to address those problems peculiar to use of thorium fuels,

.—

 
41

A feature of thorium fuel cycles in FBRs that might become very
significant in the future is related to safeguard aspects. Developing
a mixed uranium/thorium fuel cycle permits the denaturing of recycle
fissile fuels since 233U can be diluted with 238U. This flexibility
can be important if fuel feed to certain FBR power stations is limited
to about 207 enriched uranium. Either LMFBRs or GCFRs can operate
with such fuels; further, the nuclear performance of such fuels appears
to be satisfactory. The use of the mixed fuel cycle, however, leads
to some plutonium production. The use of that plutonium at restricted
sites increases the fuel-utilization characteristics possible; if the
plutonium cannot be recycled, it is more important to have a high breeding
ratio in FBRs.

Finally, the use of thorium in fast reactors leads to a fissile fuel
that is desirable for thermal reactors; this in combination with thorium
cycle use in thermal reactors helps permit the ratio of thermal-to-fast

reactors to be relatively high in a stabilized nuclear industry.
42

5. CONCLUSIONS AND RECOMMENDATIONS

A number of conclusions and recommendations are listed below,
based on the reference conditions of this study. At the same time, it
should be recognized that there are uncertainties in a number of the
economic parameters and cost estimates utilized, and the results should
not be taken out of context. Nonetheless, several variations in economic
parameters were considered, and for the specific cases investigated, the
general conclusions remained valid. Further, the results for the thermal
reactors tacitly consider that FBRs will eventually be applied and that
thermal reactors will not always be utilized to expand the nuclear economy.
In that context, thorium fuel cycles can have the advantages given. If
thermal reactors are always used to expand the economy, the use of advanced
converters has less impact on improved fuel utilization. Also, it is not
an ensured feature of thorium fuel cycles that they will be economic. The
results given in this report indicate that unless reactors such as the
HTGR are successfully developed, thorium fuel cycles in thermal reactors
will find it difficult to compete economically with the uranium cycle.
Further, the HTGR is not ensured to be an economic system under all

circumstances.

5.1 Conclusions

 

1. Development of the thorium fuel cycle is justified on the bases
of better U30g utilization, improved potential for long-term economics,
and additional flexibility with regard to fuel recycle alternatives.
Thus, introduction of the thorium fuel cycle provides additional power
generation capability in case of the delayed introduction of commercial
FBRs, or in case there is introduction of a low-gain FBR on the reference
schedule.

2. Use of LWR(Th)s rather than LWR(U)s will increase the amount of
energy generated from a given U30g resource by about 20% above the reference
value, considering substitution of thoria for urania in present type LWR
designs. Use of LWR(Th)s beginning in 1995-2000 increases the energy

generation from specified U30g resources by 12-16% relative to complete

 
43

use of LWR(U)s. However, LWR(Th) systems do not appear economic compared
to LWR(U) systems based on present commercial reactor designs even when
the U30g price is $100/1b or more.

3. If the uncertainties regarding commercial introduction of the HTGR
in the U.S. can be resolved favorably, then the HTGR appears to offer the
best combination of economics and fuel utilization with the thorium fuel
cycle., Further, possible future increases in thermal efficiency through
application of combined cycle HTGRs significantly increases economic and
fuel utilization potential.

4, The HWR(Th) system appears better suited than the LWR(Th)
system for attaining high conversion ratios. However, the capital
component of the HWR power cost appears at least as high as that of
LWRs, exclusive of the HWR requirement for heavy water, such that total
power costs of HWRs appear higher than that of LWRs for Uj0g prices less
than v$50/1b. A decrease in HWR capital costs appears important to HWR
application in the U.S. At $100/1b U30g, the HWR(Th) system is more
economic than either the LWR(Th) or LWR(U) systems. '

5. The use of HTGRs and HWRs with conversion ratios in the 0.8 to
0.9 range increases energy generation from a given U30g resource by 20
to 647%, considering introduction of these reactors by 1995-2000. (Power
growth scenarios utilized in estimating the above considered nuclear
power levels to rise to 400 to 600 GW(e) by the year 2000.)

6. Operation of thermal reactors on Pu/Th fueling appears to be
economically attractive when Pu is recovered from LWRs or enriched-
uranium HWRs. However, the use of Pu/Th fueling does not have a large
impact on fuel-utilization characteristics because of limited Pu
availability. Further, the use of Pu in this manner does not permit
it to be available for startup of FBRs. The Pu needs of FBRs under
reference introduction and growth scenarios are such that reserving
Pu for FBRs precludes large-scale use of Pu/Th fuel cycles.

7. The economic application of the thorium cycle in thermal reactors
generally requires establishment of a fuel recycle industry, particularly
for LWRs and HWRs (fuel recycle is also required for utilizing product
Pu and uranium from the uranium cycle). Without fuel recycle, the
thorium cycle can be used most effectively in HTGRs; however, recycle in

HIGRs is desirable to increase fuel-utilization performance and is
44

economically desirable when U30g costs rise above about $40/1b for the
reference conditions of this study.

8. Converter reactor operation with conversion ratios above
about 0.9 does not appear economical; the high fuel recycle costs
associated with low fuel burnups and the high fissile inventory
requirements outweight the improvement in fuel utilization achieved.

9. The discounted economic benefits from thorium cycle use in the
various reference-type reactors, and relative to LWR(U)s with Pu recycle,
vary from $1-3.8 billion for HWRs, and from $6.4-21.6 billion for HTGRs,
based on capital charge equality for LWRs, HTGRs and HWRs, economic conver-
sion ratios, estimated power growth scenarios, thorium reactor introduction
by 1995-2000, a U30g price of $100/1b, and U30g resources of 2.5-3.5
million tons. The use of thorium cycles in reference LWRs does not appear
economic relative to use of LWR(U)s.

10. The HTGR economic benefits given in (9) above are cancelled if
the unit capital costs of HTGRs are increased by $95-115/kW(e) above
those for LWRs; similarly, the HWR economic benefits are cancelled if
the relative HWR capital costs are increased by $13-18/kW(e).

11. Although the nuclear performance of the thorium fuel cycle in
FBRs is generally not as good as the uranium cycle, use of mixed cycles
in FBRs may be satisfactory and/or desirable. Also, use of metallic fuels
might be possible with thorium while not with uranium because of the
superior properties of thorium-based metal relative to uranium alloys.
Use of metallic thorium fuel improves the performance of the thorium
fuel cycle relative to use of oxide fuel; however, safety considerations
may influence the use of metallic fuel.

12. From a safeguards viewpoint, developing a mixed uranium/thorium
fuel cycle permits a "denaturing' of recycle fissile fuels, since 233y
can be diluted with 238y, This flexibility could be important if fuel

feed to certain FBR power stations were limited to ~20% enriched uranium.

Recommendations

Strong support should be given to the thorium-cycle HIGR as the best

contingency reactor in case there is a significant delay in the introduction

 
45

of LMFBRs. The emphasis should be placed on commercializing HTGRs and
introducing them on a time schedule such that they can capture a large
share of the nuclear power market by 1995-2000.
In addition to the above, continued studies and evaluations should
be carried out on the LWR(Th) and HWR(Th) systems. Since LWRs are the
most direct vehicle for thorium utilization, LWR(Th) systems should be
studied more thoroughly to be sure the conclusions of this limited study
are valid. Such studies should also evaluate LWR designs based on a more
advanced technology; in the latter case, care must be taken to consider
the influence of licensing requirements on design. Relative to heavy
water reactors, HWR(Th)s should be considered for commercial introduction
into the U.S. as a backup to the HTGR. Associated effort should include
an evaluation of the costs and schedules for commercial introduction afid
licensing of HWRs, of the capital investments required for D,0 separation
plants, and an assessment of all costs not expected to be borne by industry.
Finally, it should be noted that the above considers no limitations
on fuel use or on fuel recycle other than those associated with economic/
technical factors. If limitations on fuel recycle are imposed upon the
nuclear industry because of safeguards considerations, the use of thorium/
uranium mixed cycles in FBRs may be necessary to have a breeder economy,
and could have implications on thermal reactor fuel cycles. This situation
was examined only peripherally in this report; based on the results obtained,
it is recommended that thorium/uranium fuel cycles in FBRs be studied in

detail, along with their possible interactions with thermal reactors.
APPENDIX A

PHYSICS CONSIDERATIONS

Summary: The physics aspects of thorium fuels in both thermal and fast
neutron spectra are discussed. Higher conversion ratios (CR) are pos-
sible using the thorium fuel cycle in a thermal neutron spectrum because
of the favorable ratio of neutron captures to fissions in U-233. The

importance of this ratio in the conversion ratio can be seen in Eq. (Al):

CR=ne' - 1- losses , (Al)

where
CR = conversion ratio,
n = v/(l + a) = neutrons created per neutron destroyed (eta),
v = neutrons produced per fission (nu),
a = ratio of captures to fissions (alpha),

e' = ratio of total fissions to fissile fissions (epsilon prime).

Neutron losses to fission products and captures in higher isotopes are

also discussed.

In a fast neutron spectrum the values of alphas for U-233 and Pu-~239

are about the same. Pu-239 has a higher value of nu, and the U-238
fertile isotope has nuclear properties which yield a higher value of
epsilon prime than in the Th-232 fertile atoms. The overall result is a
higher converéion ratio for the U-Pu fuel cycle in a fast spectrum.
However, there are safety advantages associated with thorium fuel in a
fast spectrum, which are discussed in some detail. In addition, the
physical properties of thorium metal are more favorable for use as a
nuclear fuel than the physical properties of uranium metal. A fast
breeder reactor using thorium metal fuel in the core would have a lower
breeding gain than a plutonium oxide fueled core, but might require a
lower specific inventory because of the higher power density achievable
with metal fuel. This would partially compensate for the lower breeding

gain associated with thorium cycles.
A-2

Advantages of the Thorium Fuel Cycle in a Thermal Neutron Spectrum

 

Of the over 1000 naturally occurring isotopes, only two have the necessary
nuclear properties and occur in sufficient abundance to be of interest

as potential sources for augmenting our limited fissile fuel resources.
These isotopes are 232Th and 238U. Both are abundant in nature (in the
sense that if their potential could be fully realized they would suffice
to supply the world energy requirements for centuries) and both exhibit
qualitatively the same behavior when exposed to a neutron flux in the
core or blanket of a nuclear reactor. The transmutation chains for

2327h and 238U, through which they are transformed to the fissionable
isotopes 233y and 23%u, are shown in Fig. A.l. Both nuclides transform
to their fissile offspring by a single neutron capture and two successive

B decays.

From a reactor physics point of view, the relative merits of the two

isotopes as potential fuel sources depend not so much on their intrinsic .
nuclear properties (given their inherent fertility) as on the nuclear .
properties of their offspring and, to a lesser extent, on their gestation

period. It is these secondary properties which determine their ultimate

value in a power reactor economy and influence the selection of one or

the other for application in a particular reactor type.

The virtues of thorium as the fertile element in thermal reactor fuel
cycles have been reiterated by a number of investigators over the
years.1’2’3’”s5 In the final analysis, all of the arguments revolve
around the fact that 233U, the fissile daughter of 232Th, produces more
neutrons than its competitor, 239Pu, when exposed to a thermal neutron
spectrum. The difference is not large. The number of neutrons produced
per thermal (2200 m/sec) neutron absorption is n2.28 for 233y yersus
n2.11 for 23%Pu, but in the tightly regulated neutron economy of a
nuclear reactor core where control is exercised in terms of increments
of the delayed neutron fraction (from 0.657% to 0.21% depending on the
fissile component of the fuel), the additional neutrons can have a _
significant impact on the nuclear performance and economics of power .

reactor operation,

 
 

 

 

232Th + 4 NEUTRON1 FERTILE

T

2335, (274 days)
BETA DECAY

'

233 + 4 NEUTRON
90% FISSION FISSILE
10% CAPTURE

:

234 +1 NEUTRON ] FERTILE

:

235 +1 NEUTRON
80% FISSION FISSILE

20% CAPTURE

T

236 +4{ NEUTRON I PARASITE

 

 

 

 

 

 

 

 

 

 

 

 

 

 

237N p
CHEMICALLY SEPARABLE

ORNL-OWG 76-17706

 

238() +1 NEUTRON ]

2 (2.3 days)
*Ne BETA DECAY

:

239py + 1 NEUTRON
65% FISSION
35% CAPTURE

)

240py +4 NEUTRON ]

+

2¥py +1 NEUTRON
75% FISSION

25% CAPTURE

T

242p; 4+ 4 NEUTRON‘|

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

243Am
CHEMICALLY SEPARABLE

Fig. A.1l. 1Isotopic buildup in thorium and 238y systems.
A-4

The discussion which follows describes in detail those particular
physical attributes of the thorium fuel cycle which make it attractive
for implementation in thermal reactor systems. It is extracted essen-

tially verbatim from one of the review papers on the t0pic.2

Table A.l shows the important nuclear parameters of the principal fissile
isotopes available for use in nuclear power reactors. Figdre A.2 shows
the spectrum averaged n values® (plotted as n-1) for a series of binary
mixtures of 233U, 235U, and 239Py in a graphite moderator at room tem-
perature, at 573°K, and at 900°K. Figure A.2 may be taken as a good
indication of the various isotopes' potential for high conversion ratios
in thermal reactors, and it may be seen that only 233U has values of 1
appreciably larger than 2.0. (Also shown in Fig. A.2 is the "thermality"

or fraction of all neutron absorptions in fuel that occur at neutron

energies below 0.45 eV.)

The attainable conversion ratio, in a thermal reactor, depends somewhat
on the choice of moderator. The principal moderators are water, heavy
water, beryllium, beryllium oxide, and graphite. The maximum conversion
ratio for 233U in each of these moderators, allowing only for losses in
the moderator itself, is shown in Fig. A.3 as a function of the slowing-
down power, gos, per fuel atom.- (US is the free-atom scattering cross
section of the moderator, and £ is the mean logarithmic energy loss of

neutrons in collision with moderator atoms.)

The curves generally exhibit a maximum, resulting from the opposing
effects of rising n and increasing moderator loss as moderator-to-fuel
ratio increases. Losses in D,0 are very small, even with an allowance
(which is included in the curve) for 0.1l4 percent H,0 in_the D,0. The
maximum breeding ratio in H90 1is only 0.02 less than in carbon; however,
as with D,0, losses in structure may be important. Beryllium would
appear to be especially suitable as a moderator for thermal reactors;
its large (n,2n) cross section is only partly offset by a low-threshold

(n,a) reaction, yielding a net fast-effect factor of about 1.07 (for

 
Table A.1. Neutron Cross Sections (in Barns) of the Principal
Fissile Nuclides 23° U, 235U, 23°Pqy, and 2*! Pu?

(Neutron energy =0.0252 eV, velocity =2200 m/sec)

 

 

 

233U 235U 239Pu 241Pu
gab 578 +2 678 +2 1013 +4 137549
o 53142 580 +2 742 43 1007 +7
oy 47 +1 98 +1 271 +3 368+ 8
o 0.089+0.002 0.169+0.002 0.366+0.004  0.365+0.009
" 2.28440.006  2.072+0.006 2.109+0.007  2.149+0.014
b 2.487+0.007  2.423+0.007 2.880+0.009  2.934+0.012

 

* Hanna, G. C. et al 1969. A4t. Energ. Rev. 7:3-92. Figures in the referenced article
were all given to one additional significant figure.

b 0o =0740,; a=0,/s;7; v=neutrons per fission=1(1+a).
ORNL-DWG 76-47296

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

100
2
> 75
=
é
= 50
a
wl
X
= 25
0
1.3 -
293°K _
=?00°K
12 BRyax (900°K)_
\\
\—; \\
1.1
(H-1) 900°K |
z " 1 /Rr (900°K)
B 235 L MAX 1
e >A
- A~
e 0.9 o~ T
~ / (7-1)
- 23
£ *3%py
\ /
0.8 - > _>___.
0.7 /,/
0.6
10t 2 5 10° 2 5 104 2

N INy

Fig. A.2. Spectrum-averaged eta maximum theoretical breeding
ratio, and "thermality" (fraction of absorptions below 0.45 eV) as
functions of moderator-to-fuel atom ratio. Carbon moderator. Temper-
atures as indicated.

 
CONVERSION RATIO

1.4

1.3

4.2

1.1

1.0

0.9

0.8

ORNL-DWG 76-17295

 

 

Be 4+ : NN

 

 

 

 

 

D,0(293°K)

 

 

raneamen = —
——
’—

 

I~

Be (SAT. 8Li) N\

 

 

Ho
\\

[~
~

\

X

0(293°K) ]

 

(2%%Py)

[Be

 

L

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

10® 2 5 108 2 5
€I //\/23

10*

Fig. A.3. Maximum theoretical conversion ratio for various

moderators, as a function of moderator-to-fuel atom ratio.
233y and temperature is 900°K unless otherwise indicated.

Fuel is

 
A-8

Be)? or 1.04 for BeO.* Unfortunately, the Be(n,a) reaction produces
6Li, which has a neutron absorption cross section of 940 barms at 0.025.
eV, and therefore reaches saturation rather quickly — more quickly than
the fuel burns up. 1In Fig. A.3 we therefore indicate the reduced con-
version ratio that would follow saturation of the °Li. (Higher-order
gaseous products, 3H and 3He, which would result from neutron capture in
6Li, are presumed to be unimportant, as poisons, because of the long

life and mobility of the 3H.)

Fast neutron multiplication can also result from fissions or (n,2n)
reactions in 235U, 232Th, or other even-even nuclides such as 23L*U,
236U, or 240py, 1In contrast to the situation in fast-breeder reactors,
however, these reactions make only minor contributions to the overall
neutron production in thermal breeders. Fast fission in 232Th is much
less important than in 238U because the cross section above threshold is
much lower for 232Th than for 238y.

1il'h
Control of the neutron loss due to leakage is largely a matter of eco-
nomics. Leakage can be reduced by surrounding the active core by a
blanket region containing mainly the fertile material — e.g., 2327 —
the extent of the reduction depending in part on the thickness of the
blanket. Increasing the blanket thickness, we reach a point beyond
which a further increase would cost more than the value of the additional
neutrons saved. Indeed, it may be found that no blanket is economically
justifiable. In any event, as a general rule, leakage losses in a
reactor designed for minimum power cost are not likely to be less than

0.01 to 0.02 (relative to n source neutrons).

Neutron losses to the high-cross-section fission product 135%e are well
known. The xenon poison fraction — i.e., neutron absorptions in xenon

per absorption in fuel — may be related to the fuel specific power, -

 

*Based on ENDF/B-Version III cross sections for Be. Version II for other .-
nuclides.

 
A-9

S[MW(t)/kg fissiie], which is a useful generalized measure of the
neutron flux level in a reactor. Using the xenon yield for thermal-
neutron fission of 233U (0.060) and cross sections appropriate to a
graphite core at 600°C with Nc/N23 v 9000, we find for the xenon poison

fraction
P = 0.0545(0.44 + S)~1

For typical in-core inventories of fissile fuel, values of S of 1 to 3
MW(t) /kg will normally be attained, corresponding to values of P of
0.037 to 0.047. Thus, a reduction of about 0.04 in conversion ratio

will usually be associated with equilibrium concentrations of 135%e.

Following a reactor shutdown or reduction in power, the xenon poisoning
temporarily increases, passing through a maximum 10 to 12 hours after
the shutdown. The magnitude of this transient additional poison frac-
tion also depends on the fuel specific power, and is approximately 0.01,
0.04, or 0.07 for S = 1, 2, or 3 MW(t)/kg, respectively. Although the
temporary loss is not significant by itself, a reactivity reserve for
xenon override, if normally compensated by control rods, would represent

a permanent loss of neutrons.

A potentially significant neutron loss in Th fueled thermal reactors is
that due to capture in 233Pa, which is an intermediate in the breeding

Teaction

232Th(n,y)233Th~é-%n—+ 233p, _Z_ETH—’ 233y |

233pa has a thermal-neutron cross section of about 43 b and a resonance
integral of about 850 b. The loss of neutrons by absorption in 233pa
is similar to the 13%Xe loss, in that it involves a competition between
neutron capture and radioactive decay, and is roughly proportional to
fuel specific power for o(Pa)¢/A << 1. However, since absorptiofi of a

neutron by 233pg destroys a nascent 233y atom, as well as removing a
A-10

neutron that might have created yet another 233y atom, the loss to 233Pa

is double the simple ratio of absorptions in 233pa to absorptions in
233U.

We relate the 233Pa loss to specific power in a way similar to that used
for xenon. While the ratio of spectrum-averaged cross sections, 5(233Pa)/
31233U), does depend on the reactor spectrum, a value of one-third may

be taken as typical. Assuming that the conversion ratio is close to
unity, and noting that the decay comstant of 233pa is 0.0257/day, we

find that the loss in conversion ratio is given approximately by
SBR n 25(64 + S)71

with values of 0.03, 0.06, and 0.09 for S = 1, 2, and 3 MW(t)/kg, respec-
tively.

This loss may be reduced by partial segregation of the thorium and
fissile uranium so that the thorium, and hence the protactinium, ex-
periences a low neutron flux, while the fissile uranium is exposed to a
higher flux. The factors involving the specific power in the above
expression would then be multiplied by the ratio of effective flux in

the thorium to that in the fuel.

An interesting consequence of the relatively long mean life of 233Pa
(39 days) is that a significant reactivity addition can occur during a
prolonged reactor shutdown. During normal, steady-state reactor oper-
ation, the ratio of 233Pa inventory to fissile uranium inventory is
approximately $/20, where S is, again, the in-core fuel specific power
in MW(t)/kg (fissile). Thus, for S in the range 1 to 3 MW(t)/kg, an
increase of 5 to 15 percent in fuel inventory would occur, with a time
constant for approach to saturation of 39 days. While the reactivity
effect of this additional 233U would depend on its location — i.e., on
the initial degree of segregation of the fissile and fertile materials
in the reactor — the effect could be as much as 40 percent of the frac-

tional increase in fuel inventory; thus a reactivity increase Sk/k ~ 0.02

 
A-11

to 0.06 could occur. Such a reactivity increase need not be a problem,
but appropriate control devices would be required to compensate for it.
After the reactor is brought back to power, some loss of neutrons to
control poisons might be involved, while equilibrium concentrations of
fuel and protactinium are reestablished. Unfortunately, the reactivity
increase associated with 233Pa decay cannot provide xenon override
capability, since the time constants for the two processes are very

different — i.e., 39 days vs 10 hours.

One of the most important sources of neutron loss, from the standpoint
of achieving high conversion ratios in a thermal reactor, is the loss to
slowly saturating or nonsaturating fission products. In contrast to
135%e and 1%*%Sm, whose very large neutron-absorption cross sections
cause them to reach saturation very quickly, the great majority of the
fission products have cross sections which are comparable to or smaller
than that of the fuel itself. Thus, the aggregate poisoning effect of
these fission products is roughly proportional to the fractional burnup
of the fuel prior to its removal from the reactor for chemical processing.
The fission product poisoning depends also on the neutron spectrum, on
the predominant species of fuel in the reactor, on the fuel-replacement
strategy employed, and on the flux level, or fuel specific power. It is
hardly possible, therefore, to exhibit a single universal relationship
between fuel burnup and fission-product poisoning. Nonetheless we show
a few typical points in Fig. A.4 in which the fractional fuel burnup is
expressed in terms of fifa — i.e., fissions per initial fissile atom in
fresh fuel. (Note that with fuel regeneration by breeding, exposures
greater than one fifa are possible.) It may be inferred from Fig. A4
(with due allowance for the effects of other variables) that neutron
losses in the neighborhood of 0.10 (per neutron absorbed in fissile

atoms) may be expected for fuel exposures of 1 to 1.5 fifa.

Another rather important factor that tends to reduce conversion ratio in
a thermal reactor is the presence of higher isotopes of uranium, resulting

from successive neutron captures in the chain starting with 233y, The
ORNL-DWG 76-17304

 

 

 

0.14

HTGR
0.12 ®
0.10

 

//
008 / g PWR: P(X)

0.06 //

 

 

 

P FISSION PRODUCT POISONING

 

 

 

 

 

 

 

 

 

 

 

 

e
|
X =
N
/{P()m ¥ f Pl ax’
0.04 / 7 o
0.02 / /
0

 

o 0.2 0.4 0.6 08 1.0 1.2 14 i.6 1.8
X, FUEL EXPOSURE (fifa)

Fig. A.4. Fission product poisoning, excluding 135%e and !“%Sm
from 149 chain fission yield. (Neutron absorptions in fission products
per absorption in fissile fuel.)

 
A-13

reduction results in part from the weighted contribution of the lower n

of 235U and in part from the added neutron loss in 230U and 237Np.

It should also be noted that the buildup of 236U, and of any nuclides
beyond it in the chain, may be rather slow, owing to a rather small
cross section of 238U relative to that of 233y, For a fuel specific
power of 1 MW(t)/kg, the time constant for the approach of the 236y
concentration to equilibrium is something like 50 years at 0.8 plant
factor (40 equivalent full-power years). Of course a higher specific
power would produce a shorter time constant. (For this calculation, the
specific power must be based on the entire inventory of fissile uranium
chargeable to the reactor — i.e., including the out-of-pile as well as
in-pile inventories.) On the other hand, it should also be noted that
if a reactor system is started up initially with 235y, owing to a lack
of 233y for startup, then an amount of 23U much greater than the equi-
librium amount would be produced early in the life of the system, and
the equilibrium concentration would be approached from the high side.
This extra poisoning effect must be experienced somewhere in the nuclear

power complex, whether or not the extra 236U is retained in the reactor.

Quite apart from the cost factors involved, rapid chemical processing
may prove to be undesirable if the recovery of fissile material from
exposed fuel elements is incomplete. If a small fraction of the fuel is
lost during each fuel processing cycle, an effective reduction in con-
version ratio is experienced which is inversely proportional to the
discharge fuel exposure, expressed in fissions per initial fissile atom.
For example, at an exposure of one fifa (neglecting a small correction
due to radiative capture), a processing loss of 0.5 percent would give
rise to a 0.005 reduction in conversion ratio, while at 0.1 fifa the
same processing loss would lower the effective conversion ratio by 0.05.
In Fig. A.5, we see how the combined loss of conversion ratio due to
fission products and processing losses might vary with fuel discharge
exposure, for a postulated linear loss due to the fission product aggregate

(excluding 133Xe and 1“%sm).
A-14

ORNL-DWG 76-17298

 

 

   
  
 

 

 

 

 

 

 

 

 

 

0.45
o
-
<
[+
S 0.40
g PROCESSING
w
>
Z
o
o
<
ES 0.05
-
g _—-HYPOTHETICAL LOSS TO
3 FISSION PRODUCTS
o

0
0o 0.5 10 1.5

FUEL EXPOSURE (fifa)

Fig. A.5. Combined reduction in conversion ratio due to fission
products and chemical processing losses.

 
A-15

For solid fuel elements, material losses in processing and refabrication
are customarily supposed to be in the neighborhood of 1.0 percent.
Figure A.5 would suggest an optimum exposure of about 0.3 fifa for this
rate of loss; but in fact, because of fabrication and processing costs,

the economic optimum exposure would typically be much greater than this.

The differences between the 233U-232Th and 239pu-238y fuel cycles can be
explicitly quantified by considering the two flow charts given in Fig.

A.6. These charts represent the fissile and fertile transmutation

chains and neutron utilization in two idealized thermal reactor cores
optimized for the respective fuel cycles.3 The 233y-232Th system is a
heavy water moderated reactor and the 23%pu-238y system is an optimized
light water reactor. In addition, it is assumed for purposes of exposition
that no neutrons are lost to leakage or parasitic captures in structure,
moderator or fission products. The neutron balances are based on the

destruction of 100 atoms of the fissile isotope.

Consideration of these flow charts shows that for these idealized thermal
fuel cycles the 233y-232ThH system yields a conversion ratio of 1.18,
i.e., 118 atoms of 233U are produced for each 100 atoms destroyed, while
the 23%9Pu-238y cycle yields a conversion ratio of 0.99. Neither of

these values is achievable in practice because of parasitic captures and
neutron leakage, but the incremental difference between the thorium and
uranium fuel cycles carries over to actual reactors and can be exploited

by the clever nuclear designer.
Characteristics of Thorium Fuel in a Fast Neutron Spectrum

In a thermal neutron spectrum higher conversion ratios are possible with
thorium fuels with 233U as the fissile isotope because of the lower
ratio of captures to fissions in 233U compared with 235y and 23%pu.

This advantage does not exist in a fast neutron spectrum, as shown in
Figs. A.7 and A.8 from ref 5. In a fast spectrum the larger value of v
for 23%Pu dominates and results in a higher conversion ratio (higher n),
as shown in Fig. A.9. Table A.2 (ref 7) lists values of v for several

isotopes.
A-16

ORNL-DWG 76-6648
124 NEUTRONS 122 NEUTRONS

 

19

   
   
    

FISSIONS

 

233-232Th Fyel Cycle Conversion Ratio=1-18
{76 NEUTRONS 107 NEUTRONS

283
190 K
FISSIONS FISSIONS

a DECAY

  
 

 

°®pPu-28y Fuel Cycle Conversion Ratio=099

Fig. A.6. Neutron and isotopic balances for idealized Pu-U and
U-Th thermal fuel cycles.

 
A-17

ORNL-DWG 76-17705

 

 

 

 

 

 

 

 

 

 

 

 

_ 3
2
5
= 233
2 U
w
b
® : 20, J
Pu
235
U
1
0
001 040 100 1000
E(MeV)
Fig. A.7. Fission cross-section 233y, 235y, and 23%u at high
energy.
ORNL-DWG 76-17704

 

 

 

 

 

 

 

 

 

 

0.4
0.3
239 235
Pu U
0.2
T
233U

0.1

O

0.01 040 1.00

E(MeV)
Fig. A.8. Capture to fission ratio 233y, 235y, 23%y at high
energy.

 
n (£)

A-19

ORNL-DWG 76-17294

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

4.5
4.0
3.5 —
3.0 ’
o A
,/\ \ (\ ,1
2.5 [ AT —— 241 f
| 233y | I “*s |ome=- PU_.~f) /
____7_@ | 1) \"’4[ /T
’_-f ‘\ /’ I : /
2.0 .’.\qb*\\ | ' / f[l //
239, J b _ P
u | h —‘é’ss
1 u
s i 1% A
. “1
il (y
I “i
1.0 AL
|/
V
0.5
0 -2 =4 0 4 2 3 4 s 8 T
10 10 10 10 10 10 10 10 10 10
NEUTRON ENERGY (eV)
Fig. A.9. nVS energy for 233U, 235U, and 239py.
A-20

Table A.2. Neutrons Emitted per Fission (v) Produced
by 1.5 MeV Neutrons

 

 

Isotope V

Th-232 2.2
U-233 2.66
U-235 2.58
U-238 2.57
Pu-239 3.09
Pu-240 3.1
Pu-241 3.2

 

Another feature of thorium in a fast neutron spectrum is the relatively
low contribution of neutrons from fertile fissions in Th-232, compared
with U-238. An important component of the conversion ratio (breeding

ratio) is the fast fission factor, €', shown in Eq. (Al). As shown in
Fig. A.10 the fission threshold energy is considerably lower in Th-232

than in U-238, and the fission cross section is much higher in U-238 at
all neutron energies above 1 MeV. A similar plot in Ref (2) shows the
fission cross sections for Pu-240 to be significantly higher than even

U-238. Substantial quantities of Pu-240 are present in Pu-U FBR fuels.

Because of the neutronic properties described above, thorium fuels are
inferior to uranium-plutonium fuels in fast reactors, with respect to
breeding potential. In an LMFBR such as the Clinch River Breeder
Reactor, fast fissioning in U-238 and the higher fertile isotopes of Pu
contribute 17% of the power produced. ' If Th-232 replaced U-238 as the
fertile material, the fast fission contribution would drop to approxi-
mately 3%7. The contribution of U-238 to direct energy production in an
FBR is an advantage from the standpoint of resource utilization; however

this large fast fission effect has a safety drawback.

It has been shown that the Th-U fuel cycle has superior conversion
characteristics in a thermal neutron spectrum, while in a fast spectrum

the U-Pu fuel cycle has superior characteristics. For a fast spectrum

 
o FISSION (b)

10

08

0.6

04

0.2

A-21

ORNL-DWG 76-47702

 

[

 

238U

 

 

232
~n_

 

 

 

—

 

 

 

 

 

 

2

E(MeV)

6

Fig. A.10. Fission cross-section of 232Th and 238y.

10
A-22

with a median neutron energy of 0.15 Mev (typical of an LMFBR) Pu-239
has a higher breeding potential than U-233. As the average energy of
the neutron spectrum increases, the relative Pu-239 to U-233 breeding
advantage increases. This increase is due in part to the increased
contribution of U-238 fissions as the neutron energy spectrum hardens.
This characteristic of the U-Pu fuel can lead to control problems during

potential sodium voiding incidents.

Sodium voiding in an FBR core has two major effects on reactor neu-
tronics. First, the average neutron energy increases and second neutron

leakage (loss) increases.

In a large FBR fueled with 23%9Pu and 238y, the higher neutron energy
causes increased 238y fissions and a higher eta value for 23%py, These
positive reactivity effects dominate the negative leakage effect and
lead to a large increase in reactivity which can cause an unstable
control condition. In a thorium fueled reactor the lack of a signif-
icant fast fission factor causes leakage to be the dominant effect so
that the sodium void effect is negative or much less positive than in
the case of a 23%u-238y core. The positive sodium void effect for
239py-238y can occur only in large cores. The use of a thorium based
fuel would mitigate this problem and would yield improved inherent

safety for the LMFBR.

The effect of 233Pa as a neutron poison is less severe in the case of a
fast spectrum as compared to a thermal spectrum because the ratio of the
capture cross sections of 233pa and 232Th is only 1.5 in a fast spectrum
whereas this ratio is v6 in a thermal spectrum. The higher power density
in an FBR causes a higher rate of burnout of 233Pa than occurs in a

thermal reactor. This effect decreases the production rate of 233y,

The actual breeding performance of FBRs fueled with Th and 233y depends
significantly on the fuel form, specific power, and type of core cooling.
Table A.3 shows the relative breeding ratios of FBRs fueled with 233(-Th
and 23%u-U.

 
A-23

Table A.3. Breeding Ratios of 2500 MW(th) FBRs?

 

 

Power Breedi
Fuel Nuclides Fuel Composition Coolant Density Raiiong
[MW(Th)/1liter]
233
U-Th Metal Na 0.62 1.26
Oxide Na 0.39 1.16
233
U-Th Metal He 0.45 1.29
Oxide He 0.24 1.21
239 ,
Pu-U Oxide Na 0.38 1.35
Oxide He 0.24 1.44

 

#Source: B. R. Sehgal, C. Lin, J. Naser, W. B. Loewenstein,
"Thorium-Based Fuels in Fast Breeder Reactors,' Trans. Amer. Nuecl. Soc.
21: 422 (1975).

bThe.values shown are for a spherical reactor and would be smaller
for the usual cylindrical reactor.
A-24

No 239Pu-U metal fuel is shown in Table A.3 because the poor irradiation
properties of uranium-plutonium metal alloys exclude their consideration
for economical FBRs. The Th metal systems show substantial breeding
gains over Th oxide systems. The 239py-U oxide system has a better
breeding ratio than either the Th metal or oxide cases. The higher
power density considered in the Th metal system caused this reactor to
have a signficantly lower fissile inventory than the oxide fueled cores
and the lower inventory would partially compensate for the breeding

ratio difference between the thorium metal and uranium oxide cases.

 
A-25
REFERENCES

Paul R. Kasten, '"The Role of Thorium in Power Reactor Development,"

Atomic Energy Review Vol. III, No. 3, pp. 473ff 1AFA, Vienna, 1970.

Alfred M. Perry, Alvin M. Weinberg, ""Thermal Breeder Reactors,"

Annual Review of Nuclear Science, Vol 22 (1972).
E. A. Eschbach, D. E. Deonigi, "Possible Optimum Use of Thorium and
Uranium Employing Crossed Progeny Fuel Cycles," Proc. 2nd Intl.

Thorium Fuel Cycle Symp., Gatlinburg, Tenn. (May 1966).

P. R. Kasten, M. L. Tobias, "Application of the Thorium Fuel Cycle,"
Trans. Am. Nucl. Soc., ANS Mtg. San Francisco, Nov. 1975.

USAEC, The Use of Thorium in Nuclear Power Reactors, WASH 1097
(June 1969).

A. M. Perry, G. L Ragan, USAEC Document, ORNL-TM-3827 (1972).

David Okrent, '"Neutron Physics Considerations in Large Fast Reactors,"
y g

Power Reactor Technology, Vol 7, No. 2, pp. 107-137 (Spring 1964).
APPENDIX B

] THORIUM FUEL CYCLES IN LWRs

Summary: Several studies have been done which have considered the use of
thorium fuel cycles in light-water reactors to improve uranium ore utiliza-
tion. These studies are described and compared in this Appendix. Fuel cycle
cost calculations have been made for these thorium éycles using a consistent
technique (described in Appendix L) that has also been used for other
reactor concepts using thorium fuel cycles. According to the studies

cited, conversion ratios of up to 0.73 are possible using 233U02-Th02

fuel in standard LWR fuel elements, achieving burnups similar to those
achieved by LWR fuels operating on the uranium fuel cycle (about 30

MWd/kg HM). Slightly higher conversion ratios (up to about 0.79) are
possible with 233y~Th metal fuels. The reprocessing and refabrication

costs of metal fuels are not known, but it is anticipated that considerable
cost savings could be realized in fabrication if metal fuel and zircaloy
cladding were coextruded. It has been concluded that a considerable
development effort would be required to qualify the processes and product
for this concept and that the expense incurred may not be justified in

terms of the benefits received.

Conversion ratios of near unity have been calculated for metal fuels operating
to very low exposures. It is emphasized that these calculations have been
made for a standard LWR core arrangement, with no modifications except to
replace ceramic U0, with Th-U metal fuel and to decrease fuel exposures.

The fuel cycle cost calculated, assuming the same costs on a2 $/kg HM basis

as those used for ceramic fuels, was very high for this concept because

of the frequent reprocessing required.

The initial and makeup uranium inventory for all the cases considered in
this Appendix were 937% enriched in U-235. According to the studies, Pu can
be substituted for the highly enriched uranium with some economic advantage.

Nuclear performance is slightly better with uranium fissile feed.

B-1
Ore Utilization Using Thorium Fuel Cycles in LWRs

Calculations derived from several literature sources have been used as
indicators of the ore utilization capabilities of LWRs using thorium-
uranium and thorium-plutonium fuel cycles. Table B.l describes several
parameters for 1000 MW(e) PWRs operating with UO,, urania-thoria, uranium-
thorium metal, plutonia-thoria, plutonium-thorium metal, and urania-
plutonia. The table also has similar information for BWRs using U0, and

uranium—-thorium metal fuel. Several items of interest are:

l. Only the first and second cycles are included in the table. Information

on the equilibrium cycle is needed to make valid comparisons.

2. The conversion ratio for the first uranium cycle is lower than that
for the second cycle. This is contrary to what is usually reported
and is not explained in the reference. These results are probably

caused by use of inconsistent cross sections for plutonium.

3. It was assumed that the reactor is operating with annual reloading and

that the fuel is discharged after reaching 33 MWd/kg HM burnup.

4. The isotopic composition of the discharge Pu after 33 MWd/kg HM is
58.97%7 Pu-239, 21.4% Pu-240, 14.27% Pu-241, and 5.57% Pu-242.

The large amount of Pu-240 that converts to Pu-241 is the reason for
the higher conversion ratio in the plutonia-urania column, compared
with that in the U0, column. It appears that inconsistent cross

sections were used for plutonium isotopes, distorting the result.

5. The burnup character of U0y, Pu0O,, and U-Th metal fuels is compared in
Fig. B.1l. There is a very rapid reactivity change in the UO, lattice.
The large amount of Pu-240 in the Pu0; lattice helps to reduce the
reactivity swing through the full fuel cycle. The metal fuel (U-Th),

which has the highest conversion ratio of the three, shows the least

 
Table B.1l. LWR Fuel Cycle Characteristics for 1000 MW(e) Reactors

 

 

 

PWR BWR
U0, U0,~ThO, U-Th Pu0,-U0,% Pu0y-Th0,% Pu-Th% U0, U-Th
Initial fissile enrichment-w/o U-3.20 U-4,50 U~-3.91 Pu-2.37 Pu-4.48 Pu-3.71 U-2.70 U-3.67
U-0.72
Initial uniform loading U-2.740 U-3.,681 U-4.583 Pu-2.079 Pu-3.720 Pu-4.398 U-3.802 U-6.830
(MT fissile) U-0.618
Natural U (102MT) 5.028  7.189 8.951 ' 6.889 13,337
Separative Work (102KG) - 4,063 9.324 11.608 | 5.148 17,299
Conversion ratio of first fuel cycle 0.61 0.76 0.81 0.74 0.78 0.81 0.62 0.81
Makeup requirement per year U-0.459 U-0,315 U-0,176 Pu-0.253 Pu-0,310 Pu-0.302
(MT fissile) U-0, 206
Natural U (102MT) 0.824  0.615 0.344
Separative work (102KG) 1.018 0.798 0.446
Conversion ratio of second cycle 0.67 0.79 0.84 0.75 0.81 0.84

 

alues of Pu are those of Pu-239 plus Pu-241 only.

Source: C. Lin and B. Zolotar, "Thorium: An Alternative Fuel for LWRs,'" Electric Power Research Institute (EPRI)
Research Progress Report NP-2 (February 1975) p. 19.
ORNL DWG 76-17701

 

14 |- —

UO2
PuO2

1.2

 
 
   
    

8
X

10— e e e

 

 

|

' .

| |

I | |

|

- |

02 l I ' —

| | |

| I |

| | |
0.6 LI 11 L1

 

 

O 410000 20000 30000 40,000
BURNUP (MWd/MTU)

Fig. B.l. Burnup character of uranium and thorium fuels. Source:
C. Lin and B. Zolotar, "Thorium: An Alternative Fuel for LWRs," Electric
Power Research Institute (EPRI) Research Progress Report NP-2 (February
1975), p. 19.
B-5

reactivity change. The initial dip in reactivity is due to the 27-day
half life of Pa-233 decaying to U-233, The small relative reactivity
change in the U-Th fuel reduces the amount of poison control required

and therefore enhances the conversion ratio.

6. The BWRs require somewhat more fuel because of the lower power density
relative to PWRs. The same trends were noted in comparing U0, and
U-Th metal in the BWR, as had been observed for the PWR. Therefore,

further comparisons with the BWR were not considered.

7. Higher conversion ratios for plutonia-thoria than for urania-thoria
are reported in Table B.l. Additional studies by Combustion Engineering
financed by EPRI do not support this observation.! Better fuel
utilization is achieved using uranium (93% enriched) and thorium than
using plutonium and thorium.! It is judged that the CE results are
more representative of LWR performance, and these results are consistent

" with the limited calculations performed at ORNL.

A second study on the use of thorium fuel cycles in LWRs is described in
Table B.2 (ref. 2). The calculations presented in Table B.2 are for a
587 MW(e) PWR, therefore, most of the numbers cannot be directly compared
with the numbers in Table B.l, although this will be rectified later.

Several items of interest are noted from the Correa study (ref. 2).

1. Marginally higher conversion ratios are possible with metal fuel
compared with oxide fuel. Higher initial inventories were employed
with the metal fuel, with lower makeup requirements.

2. Significant improvements in conversion ratio are possible using
U-233 in place of U-235. However, without a source of U-233 (such
as from a breeder reactor), there is little opportunity to take

advantage of this improvement.
 

 

Table B.2. Fuel Cycle Characteristics for a 587 MW(e) PWR
Fuel U0, (3.3%)  Pu0,+'2tuo, 235y0,+Th0,  23300,+Th0,  235v+th  233U+Th  Pu0,+Tho,
(standard)
Av Absorption
(450 days)
Fissile 0.478 ' Q.480 0.480 0.442 0.480 0.440 0.496
Fertile 0.328 0.381 0.302 0.335 0.327 0.362 0.354
Fission Products 0.087 0.078 0.101 0.097 0.094 -0.090 0.087
Structural Materials 0.023 0.016 0.026 0.029 0.022 0.024 0.016
H»0 0.037 0.022 0.041 0.042 0.032 0.035 0.022
B-10 0.040 0.023 0.046 0.049 0.037 0.041 0.022
Initial Enrichment 3.30 3.51 4.19 3.20 3.67 2.60 4.29
;-(450 days) 1.93 1.91 2.04 2.22 2.03 2.21 1.96
€ (450 days) 1,09 1.09 1.02 1.02 1.03 1.03 1.03
Conversion Ratio 0.61 0.72 0.61 0.73 0.65 0.79 0.69
CR (450 days)
Inventory? (kg)
U-233 1405 1521
U-235 1589 333 1843 2157
Pu-239 1133 1542
Pu-241 249 335
Total 1589 1715 1843 1405 2157 1521 1877

9-4

 
 

 

Table B.2., Fuel Cycle Characteristics for a 587 MW(E) PWR (cont'd)
Fuel U0, (3.3%) Pu02+NatU02 23540,+ThO, 23310,+Tho, 2350+Th  233p+Th  PuO,+Tho,
(standard)
Consump tion
(kg/year)
U-233 -193 186 -228 183 -182
U-235 386 61 454 -9 473 ~9 -2
Pu-239 -100 166 -4 -2 437
Pu-241 =22 -5 ~4 -2 18
Total 264 222 253 177 241 174 271

 

aInventory = initial charge of fissile mass; thermal power = 1780 MW(th); thermal efficiency = 0,33 (assumed);
burnup = 900 days at full power (1.6 x 106 Mwd).

Source: Francisco Correa, "Thorium Utilization in PWRs" (MS Thesis)

(-9
B-8

Comparison of the UO, and UO0,-ThO, cases reveals a fissile require-
ment of about 22.51 kg U-235/MW(e) for 30 years of operation with
U0, and no recycle, compared with about 26,33 kg U-235/MW(e) for
U0,-ThO,. If complete recycle is assumed, the 30-year consumption
values are 16.20 and 16.Q7 kg U-235/MW(e) respectively. From this
comparison, it appears that there is little incentive to develop
235U02-Th02 as a replacement for the UO; cycle. However, the
233y0,-ThO, case with total recycle requires only 11.44 kg U-235/
MW(e) over 30 years. As indicated earlier, a supply of U-233 is
required to take advantage of this cycle. The 235Y0,-ThO, case is
made even less attractive when separative work requirements are
considered. About 1.76 times as much separative work is required
per kilogram of product for the 937% enriched fuel used in the

235U-—Th.02 cycle as that required for the 3% enriched U0, cycle.

Another study on the use of thorium fuels in LWRs was done by E. Hettergott

and R. K. Lane of General Atomic Company.3 This study is summarized in

Table B.3. The GA work in Table B.3 is compared with the EPRI study cited

earlier.

ll

“  Several conclusions can be drawn from the GA study:

More uranium ore is required by the thorium fuel cycle in LWRs than
by the uranium cycle, if recycle is not permitted. If recycle is
permitted, the thorium cycle yields slightly better ore utilization.
Without recycle, the order of preference (relative to ore utilization)
is UO,, UO,-ThO,, and U-Th metal. With recycle, the order of

preference is reversed.

The EPRI results are slightly more optimistic with respect to ore

utilization for the recycle cases and slightly less optimistic for

‘the nonrecycle cases.

Both the GA and the EPRI studies show that the amount of power
produced from a fixed-ore resource can be doubled by the application
of a thorium fuel cycle (with recycle) in LWRs, as compared with the

uranium cycle (without recycle). The advantage drops to about a 25%

increase when compared with the uranium cycle with complete recycle.

 
 

 

 

Table B.3. Regionwise Mass Flows at Equilibrium Conditions
Number of Initial Final
Assemblies Burnup Heavy Metalb U-235 U~235 Makeup Fissile Pu U-233 U-235 Fissile Pu U-233
(1 Region) (MWd/MTHM)% (kg) (kg) (kg) (kg) (kg) (kg) (kg) (kg)
Case 1 — U0, Reference
No Recycle
GAC 64 33,000 28,350 907 234 203
EPRIL 64 33,000 28,395 909 - 236 200
Recycle
GAC 64 33,000 28,485 770 537 203 257 258
EPRL 64 33,000 28,655 693 458 200 228 255
Case 2 — ThO,
No Recycle
GAC 64 34,500 27,160 1099 258 370
EPRI 64 34,500 27,140 1220 408 382
Recycle
GAC 64 34,500 27,170 657 399 370 169 441
EPRI 64 34,500 27,210 722 313 382 247 488
Case 3 — Th-Metal
No Recycle
GAC 64 25,800 36,520 1257 401 426
EPRI 64 24,100 38,880 1520 665 490
Recycle
GAC 64 25,800 36,590 741 340 426 253 537
EPRI 64 24,100 38,910 840 175 490 365 670

 

a
Megawatt-days per metric ton of heavy metal.

bOne-third of core for 1000 MW(e) reactor.
B-10

4, The advantage of the thorium fuel cycle with metal fuel over the
thorium cycle with oxide fuel is small in the GA study (about 9%).
This small increase does not justify the cost and time required
to qualify and specify metal fuel for LWRs. In the EPRI study, the
advantage is considerably larger (33%); however, the Correa results
support the GA study, therefore the conclusion is still considered

valid.

The three studies summarized in Tables B.l through B.3 are compared in
Tables B.4 and B.5. Some additional information from other studies for
the LWR uranium cycle is also presented. There is reasonable agreement
between these studies, except in a few areas that have already been
identified. The superiority of U-233 fuel over U-235 fuel is clearly
shown in these tables. In order to exploit this superiority, it will be
necessary to establish a recycle capability very early to recover and
utilize the bred U-233. The thorium cycle, using U-235, is less attractive
than the uranium cycle, assuming no recycle; and with full recycle, the
thorium cycle offers only a slight improvement, certainly not enough to
justify the time and expense to establish this cycle. Unless an external
source of U-233 (such as from a thorium-fueled FBR) can be established,
it is concluded that thorium cycles in LWRs are not attractive enough to

warrent commercialization.

As already described, the best resource utilization using a thorium fuel
cycle in LWRs is achieved using metal fuel. The studies summarized in
Table B.4 reflect a conversion ratio approaching 0.8, achieved with

233y-Th metal fuel irradiated to about 27 MWd/kg HM. Higher conversion
ratios are possible if lower exposures can be accepted. Reference (5)
presents some calculations which probably represent the limit of performance
of thorium fuel in LWRs that can be achieved without major design changes

in the reactor. Performance characteristics of the Thorium Replacement

Reactor Core (TRRC) are summarized in Table B.6.

 
B-11

Table B.4. Comparison? of Studies on LWR Fuel Cycles

 

 

Lin? Corréa® Pardued
(EPRI)  (IEA Brazil)  (BMI) aNLe  caS EPRLS
I. U0, No Recycle
(1) Core HM Inven- 85.63 81.97 92.36 94.40 85.05 85.19
tory
(2) Core Fissile 2.74 2,71 1.97 2.83 2.72 2.73
Inventory
(3) Annual Fissile 0.46 0.45 0.37 - 0.44 0.47 0.47
Consumptiond
(4) Burnup 33.00 33.27 30.00 30.50 33.00 33.00
(5) Conversion Ratio 0.61 0.61 0.61
II. Pu0,- 2tyo,
(1) Core HM Inven- 87.72 83.18
tory
(2) Core Fissile 2.70 2.92
Inventory
(3) Annual Fissile 0.46 0.38
Consumptiond
(4) Burnup 33.00 32.79
(5) Conversion Ratio 0.74 0.72
III. 235Y0,-Tho,
(1) Core HM Inven- 81.80 74.88 81.48 81.42
tory
(2) Core Fissile 3.68 3.14 3.30 3.66
Inventory
(3) Annual Fissile 0.32 0.43 0.40 0.31
Consumptiond
(4) Burnup 33.00 36.42 34.50 34.50
(5) Conversion Ratio 0.76 0.61
Iv. 233g0,-Tho,
(1) Core HM Inven- 74.81
tory
(2) Core Fissile 2.39
Inventory
(3) Annual Fissile 0.30
Consump tiond
(4) Burnup 36.49
(5) Conversion Ratio 0.73
B-12

Table B.4. Comparison® of Studies on LWR Fuel Cycle (cont'd)

 

 

Linb Correa® Pardued
(EPRI)  (IEA Brazil)  (BMI) ane  caf  Eprif
V. 235J-Th
(1) Core HM Inven- 117.21 100.10 109.56 116.64 -
tory
(2) Core Fissile 4.58 3.67 3.77 4.56
Inventory
{(3) Annual Fissile 0.18 0.41 0.34 0.18
Consumptiond
(4) Burnup 33.00 27.26 25.80 24,10
(5) Conversion Ratio - 0.81 0.65
VI. 233y-Th
(1) Core HM Inven- 99.80
tory
(2) Core Fissile 2.59
Inventory
(3) Annual Fissile 0.30
Consumptiond
(4) Burnup 27.38
(5) Conversion Ratio 0.79
VII. Pqu—ThOZ . .
(1) Core HM Inven- 83.04 74.59
tory
(2) Core Fissile 3.72 3.20
Inventory +
(3) Annual Fissile 0.31 0.46
Consumptiond
(4) Burnup 33.00 36.58
(5) Conversion Ratio 0.78 0.69
VIII. Pu0,-U0; (?35U Makeup)
(1) Core HM Inven- 94.40 85.46  85.97
tory
(2) Core Fissile 2,83 2.71 2.68
Inventory
(3) Annual Fissile 0.44 0.54 0.46
Consumptiond
(4) Burnup 30.50 33.00 33,00
(5) Conversion Ratio 0.61

 

%11 inventory values are kg/MW(e)
All fissile consumption values are kg/MW(e) yr
All burnup values are MWd/kg HM

 
B-13

bC. Lin and B. Zolotar, "Thorium: An Alternative Fuel for LWRs,'" Electric

Power Research Institute, EPRI Res. Prog. Rep. Feb. 1975, NP-2, p. 19.
cFrancisco Corréa, '"Thorium Utilization in PWRs'", MS Thesis.

W. M. Pardue et al., "A Comparison of Advanced Reactor Potentials', presented
at the ASME/ANS International Conference on Advanced Nuclear Energy Systems,
March 14-17, 1976, Pittsburgh, Pa.

°R. V. Laney et al., "A Brief Survey of Considerations Involved in Introducing
CANDU Reactors into the U.S.," unpublished ANL Report.

ffrivate communication from R. K. Lane (GA) to F. J. Homan (ORNL) dated
11 June 1976. Based on work performed by E. Hettergott (now with EXXON).

gFor accurate comparison, the annual fissile consumption numbers should be
normalized to the effected load factor assumed in the calculation. This
can be done as follows:

 

 

 

MWd(th) Day kg HM Charged 0.33 MW(e) _ MW (e)-year
kg HM year MW(th) year
Days
365 year

Consider the GA numbers for U0, (NR):

Md(e)-year _ (33.00) (85050)(0.33)

 

year 3(365) = 846
Therefore, load factor v 846 _ 0.85
1000 ’
. L 235
Fissile Consumption = 470 _ 0.56 kg U
846 MW (e)year
B-14

Table B.5. Comparison of Fissile Consumption for LWR
Fuel Cycles

 

30 yr 235U Requirements?

 

Fuel Cycle kg /MW (e)
Lin Corréa BMI ANL GA EPRI

Without Recycle

Uo, 22.43 24.59 28.93  29.93  29.93

U0,-ThO, 26.34 36.27  40.26

U-Th 27.85 41.48 50.16
With Recycle

235yo, 16.54 16.20  13.01  17.54 18.83 16.47

235y0,-Tho, 13.28 16.07 15.27 13.05

233y0,-Tho, 11.44

235y-Th 9.98 15.99 13.97 9.81

233y-~Th 11.48

 

akg 235y/MW(e) = Initial Inventory + 30 (annual makeup).

 
B-15

Table B.6. Performance Characteristics of

Thorium Replacement Reactor Core (TRRC)

 

Equilibrium exposure, MWd/ton (U + Th) 10,000
Feed enrichment, wt % fissile material 1.59
233y inventory, g/kW(e) 1.880
Breeding ratio 0.96
235y initial charge, g/kg (U + Th) 24.8
Initial uranium inventory,a tons natural U/GW(e) 589
2357 makeup, g/kg (U + Th) 1.47
Annual uranium consumption,b tons natural U/GW(e) 30
Total ore consumption,c tons natural U/GW(e) 1489

15,000
1.67
1.975
06.93
25.4
603
2.67
37
1713

20,000
1.76
2.085
0.88
26.5
630
4.18
44
1950

25,000
1.88
2.225
0.83
28.3
692
6.14
54
2312

 

Source: G. B. Zorzoli, "An Evaluation of a Near-Breeder, Low Cost, LWR Concept,"

Energia Nucleare 19(3): 151 (March 1972).
2piffusion plant discharge: 0.25% wt 235Q,

Diffusion plant discharge: 0.25% wt 235y; load factor: 80%; average over

20 years.

C
Initial inventory plus 30~year makeup.
B-16

The calculated performance characteristics tabulated in Table B.6 were

based on the following assumptions:

1. Recycle of all U and Pu produced by the Th-232 and U-233 neutron
chains, accounting for 1.57 losses during reprocessing and refabri-

cation.

2. All Pu-233 in discharged fuel decays to U-233 before reloading,
which requires at least 120 days between fuel discharge and
reloading.

3. Makeup fissile material is U-235.

The ore utilization characteristics of LWRs and CANDUs are shown in Table

B.7. A comparison of the total ore consumption values in Tables B.6 and

B.7 shows a decided advantage for the TRRC over standard LWRs and CANDUs.

Table B.7. Uranium Exploitation in LWRs and CANDUs

 

Natural Uranium Requirement, tons/GW(e)

 

Reactor System Annual Initial Inventory

Initial Inventory Consumption plus 30-Year Makeup

 

BWR 592 125 4342
PWR 406 133 - 4396
CANDU~-PHW 144 103 3234

 

Source: G. B. Zorzoli, "An Evaluation of a Near-Breeder, Low Cost, LWR
Concept," Energia Nucleare 19(3): 151 (March 1972).

 
B-17

Cost Considerations for Thorium Fuel Cycles in LWRs

Cost of thorium: The cost of thorium should not be a big contribution to

 

the fuel cycle cost. Using $100/kg thorium (corresponding to about $52/1b
ThO,) and the data from Table B.1l, the following costs are calculated:

0.26 mills/kWhr
0.377mills/kWhr

0.63 total thorium component.

initial inventory

yearly makeup

The initial inventory charge was based on the largest initial inventory
given in Table B.1l, that for metal fuel [about 115 kg of thorium for the
reactor and about 65 kg of thorium elsewhere in the fuel cycle for a total
Th inventory of 180 kg/MW(e)]. A charge of 10% year for the inital Th
inventory against yearly power production was assumed. The yearly makeup
was assumed to be all fresh thorium. Under recycle operation it would be
possible to recycle recovered thorium and largely eliminate the yearly
makeup coét. The cost of $100/kg thorium ($52/1b ThO,) is very high.
Figures B.2 and B.3 describe the cost and quantity relationships for both
uranium and thorium ore.® The basic data are about eight years old, but
are still in reasonable agreement with recently published information

for uranium ore.’ A comparison of Figs. B.2 and B.3 indicates that the
cost per quantity relationship for ThO, and U30g are similar. From

Fig. B.3, it appears that there is from 1 to 10 million tons of ThO,
available at $50/1b or less.

Fuel Cycle Costs for several of the fuel management concepts described

 

in this Appendix are tabulated in Appendix N. These costs are discussed
in the main portion of this report and compared with fuel cycle costs for

other reactor types that have been calculated on the same basis.
4/1b U30g (1968 dollars)

ORNL-DWG 76-6649

10° T Trtn r T rrrrd T T TTTH L AR P T TTTI

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

5 : ® O A i
2+ —
® USAEC-REASONABLY ASSURED
102 A . o4 O USAEC-REASONABLY ASSURED
= PLUS ESTIMATED ADDITIONAL | 3
E . 4 A USGS-REASONABLY ASSURED |
— fi A USGS-REASONABLY ASSURED |
B A o l—o—m PLUS ESTIMATED ADDITIONAL| _ )
B —— WEEKLY ENERGY REPORT | u
2 . ' ?“’" 22 MARCH 1976 &
— h——-b—o-—-l—c. —
= —
Ny _
2 —
100 | 1| 1 111l ] 1 1 LLbLLd | ! L Ll | | 1] | | 11 1L
10—1 2 5 1 OO 2 5 101 2 5 102 103 104

ESTIMATED UszOg RESERVES (10° short tons)

Fig. B.2. Uranium reserves vs price,.

 
8/Ib ThO, (1968 dollars)

ORNL-DWG 76-6650

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

3
O E—T 7177 T T TTTI1 1T T 171711 T T 11T T T TTT1T
5: ® .:
B ]
2 .
102: * o =
sE ° _
— ® USAEC —
— * O USGS ~
2r- ° REASONABLY ASSURED PLUS =
ESTIMATED ADDITIONAL
10' & *—0 ' —
51— _
— —
2}- -
100 L L LLLLLL 1L Lt L L LLLLLl L L L BN EIT
10-1. 2 5 100 2 5 "Oi 2 5 102 2 5 103 2 5 104

ESTIMATED ThO2 RESERVES (10 short tons)

Fig. B.3.

Thorium reserves vs price.

61T-4
B-20
REFERENCES FOR APPENDIX B

Private communication from Norton Shapiro and Robert Hellens (CE)

and Bal Raj Sehgal (EPRI) (April 1976).

Francisco Corréa, "Thorium Utilization in PWRs,' MS thesis.
Private communication from R. K. Lane (GA) to F. J. Homan (ORNL),
date June 11, 1976; based on work performed by E. Hettergott (now
with EXXON).

C. Lin and B. Zolotar, "Thorium: An Alternative Fuel for LWRs,"
Electric Power Research Institute (EPRI) Research Progress Report

NP-2 (February 1975), p. 19.

G. B. Zorzoli, "An Evaluation of a Near-Breeder, Low Cost, LWR
Concept,'" Energia Nucleare 19(3): 151 (March 1972).

The Use of Thorium in Nuclear Power Reactors, WASH-1097 (June 1969),
p. 30.

Weekly Energy Report, March 22, 1976.

 
“

APPENDIX C

THORIUM FUEL CYCLES IN HTGRs

Summary: The HTGR system operates most efficiently on the thorium fuel
cycle, and so relatively little development has been carried out in this
country on the uranium cycle for HTGRs. This appendix discusses modi-
fications which can be made to the reference HTGR fuel (conversion

ratio = 0.66) to increase the conversion ratio. Conversion ratios of up
to 0.82 are possible with currently achievable changes in fuel management
and thorium loading. Conversion ratios of 0.95 and higher are possible
with modifications to the fuel element which Would.permit higher volu-
metric fuel loadings than are possible under the reference design, and
with decreased fuel exposure. Fuel utilization is flearly doubled over a
30-year reactor lifetime by increasing the conversion ratio from 0.66 to
0.90, but because of the higher specific inventory required for higher
conversion ratios, the improved ore utilization is not realized for the

first 10 years of operation.

While the economic éenalties associated with high conversion ratios in .
HTGRs have not been totally evaluated, it appears that near-minimum fuel
cycle costs are achieved with a conversion ratio of 0.75 to 0.80. The
trade-offs between economics and conversion ratio are sensitive to both the
assumed costs of fuel reprocessing and refabrication and to future uranium

costs.

Fuel Utilization Considerations and Options

 

Among the types of power reactors currently in advanced stages of commercial
development, only the Light-Water Breeder Reactor (LWBR) and the High-
Temperature Gas~Cooled Reactor (HTGR) have been envisioned from inception

as utilizers of thorium.” Because of the HTGR's unique fuel design, which

consists of a mixture of thorium and uranium containing microspheres in

 

*
The Molten-Salt Breeder Reactor (MSBR) is an attractive user of the
thorium cycle, but development work on that concept has been discontinued.
c-2

individual fuel rods, a wide latitude of fissile to fertile and heavy metal
to moderator atom ratios are achievable without alteration of the basic
concept and fuel design. These parameters control, to a large extent, the
conversion ratio of the reactor system, and thus the current design HTGR is
amenable to alterations in conversion ratio over a fairly large range. For
this reason, the reactor fuel cycle can be tailored by varying the fuel
loadings and fuel exposures to meet changing economic and resource avail-

ability conditionms.

It is important to recognize that minimum short term energy costs do not
occur with maximum conserxvation of fissile resources. Increasing the con-
version ratio of an HTGR can be achieved by four basic design and operating
changes, These are (1) increase the core thorium load, (2) decrease the
core power density, (3) decrease the fuel residence time, and (4)

increase the refueling frequency. Each of these steps can carry with it

an economic penalty which may or may not be offset by the advantages of

improved fuel utilization.

The current ''reference' HTGR is designed to operate with a conversion ratio
of 0.66, not much higher than its competitors, the light water reactors,
which utilize the 238y-23%y fyel cycle to achieve conversion ratios of
approximately 0.60. Table C.1 shows the reactor parameters for the
reference HTGR which had been optimized to then current economic conditions.!
Currently projected uranium ore costs now favor a design of higher conversion
ratio, that is, CR = Q.76. Table C.2 (ref. 2) shows the increases that can
be achieved with the current fuel element design by the various strategies
listed above. The achievable conversion ratio (in the event that alternate
fuel designs are considered and the thorium loading is increased beyond what
is volumetrically possible with the current element) is also indicated. It
is noteworthy that the mined ore requirements can be reduced substantially

without resorting to undeveloped technology.

The economic penalties associated with the indicated gains in conversion
ratio are real but less easily quantified. They depend on projected eco-
nomic conditions, the scarcity and hence increased cost of uranium ore,
separative work costs, and the approximate knowledge of the cost to recycle

bred fuel and to ship and dispose of radioactive materials.

 
Cc-3

Table C.1. Optimum HTGR Fuel Cycle Parameters

under Current Economic Conditions

 

 

 

 

Core average power density, W/cm?® 8.4
Average C/Th ratio (initial core/equilibrium 214/238
recycle) .
Fuel lifetime,? years 4
Refueling interval,® years 1
Fraction of core refueled 1/4
Fertile load, kg/MW(e)
Initial core 32
Equilibrium core 29
235y requirements
Initial core, kg/MW(e) 1.40
Equilibrium annual makeup,2 kg/MW(e) 0.33
Average conversion ratio
Initial cycle 0.69
Equilibrium cycle 0.66
“At 80% average capacity factor.
Table C.2. Conversion Ratio Improvements®
Case Fissile Relative
Conversion Ratio Ore Equivalent
Reference HTIGR 0.66 1.0
25% Increase in thorium load 0.71 _ 0.85
Add semiannual fueling 0.76 0.71
Add reduced power density 0.82 0.53

(lower to 6 W/cm?)

Modified fuel rods and/or
improved fuel particles

~0.9220.95

 

*Based on 4 yr fuel residence time.
C-4

The increase in thorium loading, which is a primary means by which the con- .
version ratio can be increased, carries with it a requirement for a higher
initial fissile inventory to achieve and maintain criticality in the
reactor.” Thus, while the overall uranium utilization is increased over
the plant lifetime, the initial core and fuel recycle costs are higher,
Figure C.1l shows the cumulative uranium-235 requirements per electrical
megawatt for HIGRs with different conversion ratios,l Similar information
is given in Table C.3 in terms of ore feed requirements for three HTGRs.?
Note that while the total ore requirements are almost halved in going from
a conversion ratio of 0.66 to 0.90, the initial fissile loading is more
than doubled. The attractiveness of committing a large initial investment
to fissile inventory becomes more appealing if large price increases are

envisioned for fresh fuel feed over the life of the reactor.

Decreasing the reactor power density allows an increased conversion ratio

primarily because, for a given electrical output, a lower power density

implies a larger core volume, and hence additional volume is made available )
for fertile thorium loading. Also, neutron losses to 13°Xe and 233pa are .

reduced in a lower power density core.

The incremental costs incurred by a reduction in reactor power density
arise in part from the requirements for a larger reactor pressure vessel.
Such capital cost increases may be offset in part by a reduction in the
pumping power required to force coolant gas through the reactor core. Also,
decreasing the fuel exposure leads to an increased conversion ratio due to
the more frequent removal of fission product poisons and, hence, an im-
proved utilization of the neutrons in the reactor. Attendant costs arise

from the need for increased fuel reprocessing and refabrication.

 

Because of the constant 4 year fuel exposure, increasing the thorium
loading also decreases the fuel exposure.

 
C-5

Table C.3. Reactor Feed Requirement

 

U30g Requirement,? tons/MWe, for
Different Conversion Rates

 

 

HTGR, 0.66 HTGR, 0.82 HTGR, 0.90
Initial core 0.44 0.64 0.94
Annual reload 0.105 0.0565 0.035
40-year total 4.53 2.90 2.30

 

%At 0.3% tails enrichment, 80% annual capacity factor and
recycle of bred material.

Thus, the reduced total uranium ore requirements afforded by utilization of
higher conversion ratios can be achieved only by the acceptance of higher
capital or operating costs in other areas. It must also be emphasized that
the ore savings thus realized are achieved only after a considerable period
of operation. (Compare the cumulative ore demands of the 0.66 conversion
ratio case and the 0.90 conversion ratio case for the first 10 years of
operation in Fig. C.l.) As shown, the mined ore requirements for the high

conversion systems are higher during the early years of operation.
HTGR Near Breeders

 

The interest in thermal break-even breeder systems as a contingency
position in the event LMFBRs are not commercialized on the schedule now
assumed has resulted in several studies conducted at General Atomic
Company3’L+ on near-breeder HTGRs. A near-breeder HTGR system would be
similar to the Light-Water Breeder Reactor (LWBR) system described in
Appendix F, in that a 2337 jnventory is produced in a conventional HTGR
to fuel a near-breeder machine. Several pre~breeder alternatives have
been considered. They are summarized in Table C.4, from which several

observations can be made:

1. The best pre-breeders (i.e,, the largest ratio of 233y produced
to 235y consumed) have low power densities and intermediate thorium loadings
(C/Th ratios of 150 to 170). Low power densities result in higher capital
costs.

2. If fuel from the pre-breeders is not reprocessed for the first 11
years of operation, the gross U-235 requirement is such that there is little
influence of power density on optimum U-233 production. However, if the
"residual" U-235 can be recycled in the pre-breeders, there is strong incen-
tive for lower power densities (independent of the influence of higher
capital costs).

3. The results shown in Table C.4 are in agreement with German calcu-
lations associated with near-breeder systems in pebble bed HTRs (ref. 5).

A comparison of several HIGR near breeders is shown in Table C.5. Also
shown are German calculations for near-breeder pebble bed HTRs.® The GA
and German calculations are in good agreement; the differences in net

U-233 makeup can be explained by differences in the calculations:

1. The GA calculations assumed 98% enriched U-233 from GCFBR blankets
and the German calculations assumed self-generated U-233, The GA calcu-
lations for the case corresponding most closely to the German near-breeder
case were repeated using a fuel composition of 71% U-233, 207% U-234,

7% U-235, 2% U-236 and 0% U-238, and the results are shown in the comparison
given in Table C.6. The use of 98% U-233 feed in the initial core reduces
the initial core requirement by 228 kg fissile relative to using HTGR bred
uranium. Eventually, the makeup requirement for the case with GCFBR U-233
will become essentially equal to that for the HTGR U-233 case, since most

of the total charged material will be self-generated U-233 of equilibrium
composition. This equilibrium situation is not reached in the first ten
years when starting with 98% U-233, The makeup is higher in the earlier
years to compensate for the buildup of U-234.

 
Table C.4. Pre-Breeder Characteristics*

 

 

 

 

 

 

 

Case  W/CC  C/Th  Gross U-235  Net U-235  U-2331 U-233 U-233
kg kg kg Net U-235 Gross U-235

1 6 150 10520 8532 4409 0.516 0.419
2 7 240 7620 7314 2668 0.365 0.350
3 7 190 8480 7695 3298 0.429 0.389
4 7 150 10090 8470 4038 0.477 0.400
5 9 190 8577 7970 2915 0.366 0.340
6 8.3 240 7695 7472 2487 0.333 0.323
7 8.4 140 10360 8714 3878 0.445 0.374
8 4.5 365 7042 7061 2338 0.331 0.332
9 10 80 28097 14250 5838 0.410 0.21
10 6 240 7883 7330 3009. 0.410 0.382
11 4 240 8519 7396 3844 0.519 0.451
12 4 140 14026 9428 5734 0.608 0.409
13 9 215 8000 7664 2583 0.337 0.323

L=D

 

*
Kilograms for 11 years of operation at 1000 MW(e).
About 4500 kg U-233 required for starting CR = 0.97 breeder.
Table C.5. Preliminary Comparisons for HTGR Near Breeders

 

HTGRS* Near Breeders

 

C/Th = 70 C/Th = 90

C/Th = 110 C/Th = 110

Pebble Bed™

Feeder Breeder
C/Th = 205 C/Th = 125
C/HM = 198 C/HM = 110

 

Fuel Life 4 yr 4 yr
Power Density 5 w/ce 5 w/ce
CR 1.0 0.97
ICT Inventory,

kg/1000 MW (e) 4907 3545
Ic" + Reload 1

Inventory kg/

1000 MwW(e) 6292 4569
Net Makeup

(10th year)
kg 233/1000 M-

year
17 losses < 69 < 75
2.8%Z losses < 95 < 93

4 yr
5 w/ce
0.94

2822

3690

< 103
< 119

2 yr
5 w/cc
0.97

2822

4328

< 124
< 146

1.2 yr 2.2 yr

5 w/ce 5 w/ce

0.76 0.97

1414 2710 U-233
+284 U-235

235 37

 

*
HTGRs: 3% Ak leakage; GCFR U-233 feed.
Pebble bed: 27 Ak leakage; pebble bred U-233 feed.

t

IC = Initial core; IC inventory and reload requirements for HTGRs and pebble bed

reactors are based on 807 yearly capacity factors.

+
Cases not truly equilibrated at 10th year.

10th year net requirements.

Equilibrium values are 1/3 to 1/2 the

8-0

 
‘ .

C-9

Table C.6. Comparison of GAC and German Calculation of
Near Breeder Using HTGR Bred Uranium Feed

 

 

 

GAC HTGR German Pebble-Bed
C/Th = 110 C/Th = 125
C/HM = 110
Fuel Life 2 years v2,2 years
Power Density 5 w/cc 5 w/ce
CR 0.97 0.97
Leakage, % Ak 3% 2%
IC Inventory 2776 U-233 2710 (U-233)
274 U-235 +284 (U-235)
kg fissile/1000 MW(e) 3050 2994
Reload 1
kg fissile/1000 MW(e) 1882 Not given
Net Makeup
kg fissile/1000 MW(e),
with 17 losses and 31 37
80% yearly capacity
factors

 

2, The GA calculations assumed 3% Ak/k in leakage and the German

calculation assumed 2% A»/k. This effect is small compared to the initial
U-233 enrichment. A 1% Ak/k leakage difference is estimated to change the
core fissile inventory by less than 50 kg.

From the studies summarized in Table C.4, coupled with capital cost estimates
for the pre-breeder HTGR, an overall optimized pre-breeder HTGR has been
selectéd.6 This reactor has a specific power of 5 w/cc and a C/Th ratio of
170 (C/Th = 150 for reloads). For best near-term ore utilization, this pre-
breeder would be coupled with a near-breeder HTGR with a conversion ratio of
0.97. While the near-breeder HTGR continues to require makeup fuel, it re-
quires a considerably lower initial fissile inventory than the HTGR break-

even breeder (see Table C.5).
C-10

Economic Considerations

 

The decision to implement steps toward higher conversion ratio systems are
dependent on economic considerations which, in turn, rest on speculative
economic projections. For example, Fig. C.2 shows the results of a study7
to determine an optimum HTGR conversion ratio based on projected real-cost
increases in U30g of 2% and 6% per year. The low base price of $26/1b Uj30g
was assumed in the calculations presented in Fig. C.2. Other factors being
equal, the higher future uranium costs favor the higher conversion ratio

system,

In summary, the HTGR is a technologically developed reactor system that has
the capacity to extend the electrical power generation obtainable from a |
given fissile uranium resource considerably beyond the value associated with
light-water reactors operating on the uranium-plutonium cycle. Practical
conversion ratios of up to about 0.82 are possible for this system with
present technology. Conversion ratios above 0.9 can be obtained, based on
physics considerations. The actual impact that the system may have on the
nuclear resource picture depends first on its commercial acceptance; second,
on its introduction rate vis-—-a-vis light-water reactors; and third, on the
economic picture existing at the time of its introduction. These factors

are considered in Appendix G and Appendix N.

 
FISSILE REQUIREMENT (Kg 235U/Mwe)

16

14

12

10

c-11

ORNL-DWG 76-17297

 

I i }
BASED ON 80% ANNUAL CAPACITY FACTOR

 

CR= 066

 

N
O
N
\“’

 

 

 

 

\\\\\?8\\\

 

 

 

 

 

 

 

 

10 20 30 40
YEAR

Fig. C.l. Cumulative uranium and ore requirements.

/A

_ O.QO/J 20

50

075 — 3.0

1.0

URANIUM FEED (ST Uz0Og/MWe)
FUEL CYCLE COSTS (mills/kWh)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

2.2 18
6% ORE. SCARCITY
2.4 P 17
C/Th=400
=
X :
20 S 16
5W/cc E
10 W/cce =
N
|._.
3
19 S5
C/Th=150 w
—_
O
>
O
18 314
w
A T
7.5 W/cc
C/Th=200
17 1.3
16 1.2
o7 08 09

CONVERSION RATIO

Estimated HTGR fuel cycle costs for two assumed uranium scarcity values.

Fig. C.2.

v

ORNL-DWG 76-6647

 

T |
2% ORE SCARCITY

 

 

 

m
/

 

C/Th=150
/ 8W/cc

 

 

 

C/Th=200

 

 

 

 

 

o7 08
CONVERSION RATIO

09

¢T-D

 
1.

c-13

REFERENCES FOR APPENDIX C

L. L. Bennett and R. K. Lane, "Fuel Management Flexibilities of
Th-233U Cycle," Proceedings of American Nuclear Society Topical
Meeting on Gas-Cooled Reactors, Gatlinburg, Tenn. (May 1974).

R. H. Brogli, M. L. Hays, R. K. Lane, and R. F. Turner, The High
Conversion HTGR for Resource Conservation, GA-Al3606 (October 1975).

Private communication, R. K. Lane (General Atomic) to F. J. Homan
(0ak Ridge National Laboratory), June 11, 1976.

R. K. Lane, R. Brogli, R. F. Turner, and N. Koo, High Conversion
and Near-Breeder HTGR, GA-Al4029 (July 1976).

E, Teuchert and J. J. Rutten, "Near Breeding Thorium Fuel Cycle in
the Pebble Bed HTR," Proceedings of the IAEA-OECD (NEA) International
Symposium on Gas-Cooled Reactors With Emphasis on Advanced Systems,
Julich, Germany, Oct. 13-17, 1975.

Private communication, R. K. Lane (General Atomic) to F. J. Homan
(0ak Ridge National Laboratory), June 21, 1976.

R. H, Brogli (General Atomic), personal communication to J. D. Jenkins
(Oak Ridge National Laboratory).
APPENDIX D

THORIUM FUEL CYCLE IN HWRs (CANDU)

Summary: The fuel utilization characteristics and fuel cycle cost for

CANDU reactors are compared in this Appendix with LWRs. CANDU reactors

are considered operating with no recycle, with plutonium recycle, and

operating with a thorium—uranium fuel cycle. Several conclusions were

reached:

1.

Considerable flexibility exists with the CANDU system, and con-
version ratios of up to 1.0 can be achieved using the thorium fuel
cycle. An economic penalty is associated with higher conversion

ratios since lower specific power and lower burnups are required.

CANDUs operating on the thorium fuel cycle with a conversion ratio
of 0.9 require 0.43 times as much U30g as LWRs operating on the
uranium cycle with plutonium recycle. Numerous other comparisons
are made between CANDUs and LWRs operating with both uranium and
thorium cycles, with and without recycle. The CANDU is superior to

the LWR in fuel utilization.

When actual yearly ore requirements are considered (instead of the
30 year reactor commitment described above), it requires up to 15
years before the cumulative ore requirements of a reactor operating
on the thorium fuel cycle with high conversion ratio are less than
the cumulative requirements for a CANDU reactor operating with no
recycle. This is due to the high specific inventory required to

achieve high conversion ratios.

The lowest fuel cycle costs were achieved with CANDUs operating on
the thorium fuel cycle with conversion ratios in the range of 0.85

to 0.90 based on available information.
D-2

Operating Characteristics and Fuel Utilization in CANDU Reactors

 

This portion of the report presents the results of a very brief survey
of the issues, the advantages and disadvantages, that would require
consideration in any future decision to utilize CANDU reactors in the
U.S., specifically those fueling options that will utilize thorium.
Factors considered and discussed here are the relative performance
characteristics of the fueling options, the inherent U30g fuel utili-
zation economics, and the relative fuel cycle costs. The U.S. designed
PWR, both with and without uranium and plutonium recycle, has been taken

as the standard for comparison.

The fuel cycle characteristics of the standard CANDU (but uprated in
specific power from the 19 kWth/kg HM of the Canadian Pickering Plant)
both with and without plutonium recycle, and for two versions of the
thorium-fueled CANDU, are presented in Table D.l. Two basic fueling

options are considered for the use of thorium in the CANDU concept.

One option (when the conversion ratio is less than 1.0) utilizes plu-
tonium makeup for the equilibrium refueiing cycle and the other utilizes
highly enriched (v93%) U-235 as the makeup fuel. Both options will
require substantial amounts of highly enriched U-235 for startup of the
systems. Both options appear to be capable of being designed to operate

with conversion ratios of 1.0.

Some of the potential tradeoff associated with high fuel utilization in
CANDU reactors are listed in Table D.2. The use of thorium can improve
the conversion ratio in any thermal reactor type, but the neutron economy
of D,0 moderation and cooling allows more potential for high conversion

ratio,

Increasing the specific power reduces core size and D,0 inventory. Fuel
utilization is decreased as evidenced by greater fissile makeup and

lower conversion ratio to maintain the same burnup. Optimization for

 
 

 

 

 

Table D.1. Performance Characteristics of Fueling Options for CANDU Reactors
CANDU
No Pu CANDU-Th (Pu Makeup, from Argonne Study) CANDU-Th (U-235 Makeup)**
Recycle Recycle* A B C D E F 1 2 3 4 5
Specific Power, kWth/kg HM 26 26 29 29 29 29 29 29 38.4 38.4 38.4 25.6 25.6
Inventory, MT/GWe 128 128 115 115 115 115 115 115 86.8 86.8 86.8 130.2 130.2
Discharge Burnup, Mwd/kg HM 7.5 18 10 20 25 33 40 44 15 27 44 8.5 27
Fuel Residence Time, Years 1.0 2.37 1.2 2.4 3.0 3.9 4.7 5.2 1.52 2.74 4.47 1.29 4,11
Equilibrium Cycle Loading***
MT HM/GWe-yr, U 127.7 31.5 0 18.8 24.5 34,2 40.7 45.6
U-Pu-0, 40.1 0 1t 2t gt 6t gt
(Th + U-233)-02 95.7 40.8 30.9 21.2 16.3 14.0 64.9 36.0 22.1 114.5 36.0
Fissile Enrichment % U-235 0.711 0.71(U)
Fissile Pu, wt % 1.26(Pu}
Equilibrium Cycle Discharge,
wt %, U~-235 0.17 0.17
Fissile Pu, wt % 0.27 0.33
Requirement, ST U30g/GWe-yr**x*
0.2 Tails 168 94 0 24.7 32.2 44.9 53.4 59.9 31.2 40,2 52.6 0 21.9
0.3 Tails 38.8 49.9 65.4 0 27.2
Sep. Work, MT SWU/GWe-yr***
0.2 0 0 3.1 40.1 52.5 0 21.9
0.3 0 o 26.4 34.0 44.6 0 18.6
Conversion Ratio 0.74 0.74 1.0 0.96 0.93 0.90 0.87 0.85 0.90 0.87 0.82 1.0 0.93
¢, Capture to Fission Ratio 0.2 0.32 0.10 0.12 0.13 0.15 0.18 0.19 0.12 0.12 0.12 0.12 0.12
BEquilibrium Cycle Fissile Loading
g U=-235/kg HM 7.11 3.13 0 1.9 4.4 9.4 0 2.4
g Fissile Pu/kg HM 0 7.06 o 1 2 4 6 8
g Recycle U Fissile/kg HM 0 o 18 18 18 18 i8 18 18 V18 18 17.2 16.4
Total Fissile, g/kg HM 7.11 10.19 18 19 20 22 24 26 20 22 27 17.2 18.8
Average Fissile, g/kg HM 7 18 18.5 19 20 21 22 19 20 23 17.2 17.6
*See Annex D.2 for plutonium recycle basis.
**personal communication from A. M. Perry, IEA, to E. H. Gift, March 1, 1976.

***At 80% locad factor.
tFissile Pu only, g/kg HM
Table D.2., CANDU-Th Cycle Consideration

 

Ranges Optimum Trend for

 

Pickering Standard Considered Major Effect on

Station PHW-NU* for Capital Cost and Capital U30g Cost
Data Data Th Cycle Fuel Utilization Intensive Intensive

Fabrication =
Refabrication
Cost Intensive

 

Specific 19 26 1l6-38 Increased specific High** Low
Power (28 pins) (37 pins) Power :
(kWth/kg HM) 1. Reduced D,0

inventory

2. Poor fuel utiliza-

tion - increased

fissile makeup or

reduced burnup

Lattice 28.6 28.6 28.6-22.9 Reduced lattice pitch: Low High
Pitch (cm) 1. Reduced D,0 '
inventory
2. Poor fuel utiliza-
tion - increased
fissile makeup or
reduced burnup

Burnup 7.5 7.5 10.0 up Increased burnup: - Low
(MWd/kg HM) 1. Poor conversion
ratio
2. Increased fissile
makeup

Coolant PHW PHW PHW BLW & OCR: Reduced BLW -
BLW D,0 inventory or
OCR OCR: Higher thermal OCR
efficiency and higher
specific power limit

Medium

High

Medium

 

*CANDU-PHW, natural uranium fueled.
**For example, a "High" specific power is the optimum trend if capital costs are dominant.

7-d

 
D-5

high specific power is desirable if capital costs are high, and undesirable

if U30g costs are high.

Reducing the lattice pitch tends to reduce the D,0 inventory, but the
overwhelming effect is to reduce the fuel utilization. Thus optimization
would probably maintain the present lattice pitch, especially when

U30g and fabrication costs are high.

Increasing the burnup of CANDU-Th options from 10 MWd/kg HM rapidly
reduces the achievable conversion ratio, thus increasing fissile makeup
requirements. If fabrication and reprocessing costs are high, the fuel

cycle economics would favor high burnup over improved fuel utilization.

Changing the coolant from D,0 to either light water or organic substan-
tially lowers the D,0 inventory, thus substantially lowering capital
costs. The conversion ratio is lowered as a result of increased neutron

losses in the coolant.

Utilizing the data of Table D.1 and the derived data shown in Table D.3,
estimates have been made of the natural uranium fuel needs of CANDUs and
PWRs, with and without plutonium recycle and of CANDUs fueled with

thorium.

Table D.4 shows, on an individual reactor basis, the U30g requirements
relative to a CANDU having no recycle. Thus the PWR at no recycle will
require 307 more U30g than the CANDU with no recycle; for uranium recycle
alone, the PWR and CANDU U30g needs are nearly equal; with uranium and
plutonium recycle the PWRs' fuel needs are 80% of those of the CANDU

with no recycle.

Recycle of Pu in the CANDU system leads to a 407 fuel saving (essentially
the same percentage fuel saving that plutonium + uranium recycle yields
for the PWR). The use of thorium with either fissile plutonium or

U-235 makeup leads to fuel savings of 70-757 for conversion ratios near
1.0. Even for a conversion ratio near 0.8, fuel savings of 60% could be

expected.
Table D.3. Data for Use in Mined Fuel Needs Estimates

 

 

 

30-Year
Mined Fuel
Average 100% Need at
Specific Inventory* Thermal Capture Load Factor 0.8 Load
ST Natural kg Efficiency, Conversion to Fission Doubling Factor,
Reactor Description U30g/GWe Fissile/GWe % Ratio Ratio, O Time = D** ST UiQg
CANDU, No Recycle l68*** 0.30 0.74 0.20 -1.0 5208
CANDU, Pu Recycle 303 1.456 0.30 0.74 0.32 -2.58 3100
CANDU, Th-Pu Fueled
CR = 1.0 1360 5.38 0.30 1.0 0.10 o 1360
CR = 0.96 970 3.83 0.30 0.96 0.12 -70.27 1301
CR = 0.93 907 3.59 0.30 0.93 0.13 -37.31 1490
CR = 0.90 870 3.43 0.30 0.90 0.15 -24.52 1722
CR = 0.87 860 3.40 0.30 0.87 0.18 -18.22 1993
CR = 0.85 875 3.46 0.30 0.85 0.19 -15.93 2193
CaNDU, Th-U-235 Fueled
CR = 0.9 990 3.9 0.30 0.9 0.11 -28.88 1813
CR = 0.87 760 3.0 0.30 0.87 0.12 -16.94 1837
CR = 0.82 735 2.9 0.30 0.82 0.13 -11.72 2240
CR = 1.0 1470 5.8 0.30 1.0 0.10 o 1470
CR = 0.93 940 3.7 0.30 0.93 0.12 -38.79 1522
PWR, No Recycle 495 0.33 0.6 0.2 -1.88 6800
PWR, Uranium Recycle Only 495 0.33 0.6 0.2 -2.42 5410
PWR, Plutonium Recycle 495 0.33 0.6 -3.06 4370

 

*Based on 700 days ex-core inventory.

2.74 » (Thermal Efficiency) (Specific Inventory)
(Conversion Ratio - 1.0} (1 + o) ({(Reactor Load Factor)
***Tnitial core and annual relcad at 0.8 load factor.

 

**Doubling time is defined as:

 
Table D.4. Relative 30-Year U304 Requirements
(at 0.8 load factor)

 

Relative U30g Need

PWR, No Recycle : 1.31
PWR, Uranium Recycle 1.04
PWR, Plutonium and Uranium Recycle 0.82
CANDU, No Recycle 1.0
CANDU, Pu Recycle 0.60

CANDU, Th-Pu Fueled

CR
CR
CR
CR
CR
CR

[}

1.0

0.96
0.93
0.90
0.87
0.85

0.26
0.25%*
0.29
0.33
0.38
0.42

CANDU, Th-U-235 Fueled

CR

CR
CR
CR
CR

0.9
0.87
0.82
1.0
0.93

0.35
0.35
0.43
0.28
0.29

 

*small variations from expected values result from approximations in
the data of Table 2. '
D-8

Although these fuel savings are real they cannot be realized at the
beginning of the reactor lifetime. In fact, for the thorium fueled
reactor having a conversion ratio of 1.0, the entire fuel requirement is
essentially required at the beginning of reactor life. Figure D.l shows
the U30g usage pattern over a 30-year reactor life (at 0.8 load factor)
of selected 1000 MWe CANDU reactor options. Approximately seven years
are required before the mined fuel needs of a CANDU, no recycle reactor
surpass those of a CANDU reactor on the thorium cycle with a conversion
ratio equal to 1.0. Similarly about 12 years are required for the
plutonium recycle option to exceed the thorium cycle having a conversion

ratio equal to 1.0.

To gain a better feeling for the dynamics of ore utilization of different
concepts and reactor options the following scenario can be employed.
Assume, for comparative purposes, that all of these concepts and reactor
options are available now and that a country is going to choose one
concept and meet all its power needs for 50 years with this one concept.
Also, it might be assumed that fuel utilization is of major importance.
Since the actual power growth cannot be known with any certainty, the
representative power growths shown in Figure D.2 have been chosen.
(Growth B [5 GWe/yr] of the figure is approximately one-fifth of the
growth rate of the 1975 low ERDA growth projection.) The growth rates
A, B, and C of Figure D.2 lead to total nuclear capacities at the end of
50 years of 370, 250, and 170 GWe, respectively.

Figure D.3 considers growth rate A (370 GWe in 50 years) and compares
the relative Uj30g reqdirements of the PWR and the CANDU reactors (no
thorium concepts). First it is apparent that over the 50-year time
period the total cumulative requirements are relatively the same as
those for the individual reactors (as discussed in Table D.4). It is of
interest to note that although the 50-year uranium requirements for the
PWR with plutonium recycle are only about 857% of those for the CANDU, no
recycle; for the first 17 to 18 years of the campaign the U30g require-

ments for the CANDU, no recycle are slightly lower.

 
D-9

- ORNL-DWG 76-17303

- 5000 //
CANDU, NATURAL U \7/
,
4000 //

CANDU, Pu RECYCLE /

// /

CANDU, Th-233y
UEL. CR=0.82

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

CUMULATIVE UsOg REQUIREMENT (ST U30s)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

. 2000 /
/ / \<CANDU, Th-Pu
/ CANDU, Th-2%U FUEL. CR=0.85
, iéé}gf”;fiFUEL.CR=Lo I
71 Z=
— | W_cANDU, Th-Pu FUEL
—*""”(::j CR=1.0
1000 i
/ CANDU, Th-Pu FUEL
CR= 093
0
0 5 10 15 20 25 30 35

REACTOR LIFE (years)

Fig. D.1l. U30g usage pattern for CANDU fueling options.
D-10

ORNL-DWG 76-17302

 

500 l I I

Assumptions

— For Growth A,l-"-.":t""|
For Growth B, P=5¢
For Growth C, P=5¢09

 

 

400

 

 

 

 

NUCLEAR CAPACITY (GWe)

 

 

//
o //
/ | >
200 4
/
//; /c’/
T
100 /Z/

yd
~

 

 

 

 

 

 

 

 

 

 

 

 

 

pd
T
0 /
0 5 10 15 20 25 30 35 40 45 50

CAMPAIGN YEAR END

Fig. D.2. Assumed CANDU nuclear growth schedules for estimating
natural uranium utilization.

 
D-11

ORNL-DWG 76-17301

 

 

2.0 T :
ASSUMPTIONS l
18 NUCLEAR GROWTH RATE= 5 !.16We
PWR, NO RECYCLE /
16 ,

 

PWR, U RECYCLE ON LY\ //
14

) CANDU, NO RECYCLE-\\ZI /7 /
| SN/

1.0 S

 

 

 

 

 

 

 

 

CUMULATIVE U308 REQUIRED (ST U30®x 1076)

 

 

 

 

 

 

 

 

 

 

 

 

. . 0.8 //V/ "4
. / /|
06
/ /|
/ / 7
04 7 -
I
on 7 U RECYCLE
" CANDU, Pu RECYCLE
0
O 5 410 15 20 25 30 35 40 45 50

CAMPAIGN YEAR END

Fig. D.3. Cumulative U30g requirements for PWR and CANDU with and
without plutonium recycle.
D-12

are compared with those of three CANDU thorium options for growth rate A -

In Figure D.4 the cumulative U30g requirements for the CANDU, no recycle,

(of Figure D.2). Figure D.5 presents a similar comparison for growth

rate C. These figures show the effect of the high initial fuel loadings

required by the thorium fueled reactor options. 1In Table D.5 these .
results are compared with those previously reported in Table D.4 for the

individual 30-year reactor requirements. Table D.5 shows that, as the

 

growth rate increases, a longer time is required before potential fuel

gains of an advanced converter are realized.

This effect shows up also in the time required before the fuel require-

ments of an advanced converter are less than those of a less neutron

efficient concept. This is illustrated in Table D.6 which lists the

time in years before the fuel requirement of a high conversion ratio,

thorium-fueled CANDU reactor is less than the fuel requirements of all

lower conversion ratio concepts. For example, for power growth A and

the thorium CANDU having a conversion ratio equal to 1.0, approximately .
15 years are required before its fuel requirements are less than those .
of the CANDU, no recyle, 23 years before they are less than CANDU, Pu

recycle, 19 years for thorium CANDU, CR = 0.85 and 38 years for thorium -
CANDU, CR = 0.93.

In addition to the cumulative U30g requirement, it is of interest to
consider the amount of U30g that is committed for the lifetime of a
nuclear power growth campaign. At any particular time, in an expanding
nuclear growth campaign, the committed U30g is much greater than the
amount actually required. Table D.7 compares the committed and cumula-

tive U30g requirements for 50-year growth for the power growths A and C.

For power growth A, the cumulative requirement at the end of 50 years
for the PWRs and CANDUs utilizing plutonium and uranium fuels only is

one~-half of that actually committed.
6)

Cumulative U, Op Required (ST usoex 10

D-13

ORNL-DWG 76-17299

 

10 T T 1

Assumptions

— 4. Power Growth Rate = 5¢0-9

2. Individual reactor U:,’O8 require-
8 I— ments as given in Table 3.

: CANDU, No Recycl
, ( ’//
6 N\l /
5 ' ’>/
GANDU,PU ?;
4 ™~
Guorrey ]~
3 ]
/

=

 

 

 

 

 

 

 

 

 

 

 

 

    

 

CANDU, Th-Pu
CR=1.0

, /

o%

0 S 10 15 20 25 30 35 40 45 50

 

 

 

 

 

 

 

 

 

 

 

 

 

CAMPAIGN YEAR END

Fig. D.4. Cumulative U30g requirement for CANDU reactor options.
6y

Cumulative U O, Required (ST Uy0x 10

ORNL-DWG 76-17300

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

10 ‘ l ! I
CANDU,
Assumptions No Recycle\ /
9 [~ 4. Power Growth Rate = 5t {4 CANDU
2. Individual reactor Uy Ogrequire- / Pu Roc’ycle/
g |~ ments as given in Table 3 / 7
, / // ,
6 / £ CANDU, —
/ / Th-Pu
S / /
/ /V -~
4 / / '
. / L
/ //%GANDU,TI} -Pu
CR=4.0
2 / / i
Z —CANDU, Th-Pu
CR=0.93
1 /””,r'
0 .A%
0 5 90 15 20 25 30 35 40 45 50

CAMPAIGN YEAR END

Fig. D.5. Cumulative U30g requirement for CANDU reactor options.
D-15

Table D.5. Effect of Power Growth Rate of Cumulative (50-Year)
U30g Requirements

 

 

 

Individual Nuclear Nuclear
Reactor Type Reactor Growth A¥ Growth C**
CANDU, No Recycle 1.0 1.0 1.0
CANDU, Pu Recycle 0.595 0.610 0.605
CANDU, Th-Pu Feed _
CR = 0.85 0.423 0.524 0.506
CR = 0.93 0.287 0.356 0.337
CR = 1.0 0.261 0.326 0.296
1.1 . .
*Nuclear growth A = 5 t , GWe (t = years). 50-~year capacity is
370 GWe. 0.9
**Nuclear growth C = 5 t °~, GWe. 50-year capacity is 169 GWe.
Table D.6. Time Required to Achieve Breakeven Fuel Needs

 

 

No Recycle Pu Recycle CR = 0.85 CR = 0.93

 

A. Individual Reactor, 30-Year Life

 

 

 

CANDU, No Recycle o
CANDU, Pu Recycle 1 0
CANDU, Th-Pu
CR = 0.85 6 11 0
CR = 0.93 5 8 2 0
CR = 1.0 7 11 11 28
B. Power Growth Rate A
CANDU, No Recycle 0
CANDU, Pu Recycle 1 0
CR = 0.85 12.5 30 0
CR = 0.93 10 18 1 0
CR = 1.0 15 23 19 38
C. Power Growth Rate C
CANDU, No Recycle 0
CANDU, Pu Recycle 1 0
CANDU, Th-Pu
CR = 0.85 11 27 0
CR = 0.93 9 17 1 0
CR= 1.0 13 22 17 35

9T-d

 

 
Table D.7. Comparison of Committed and Cumulative
U30g Requirements for a 50-Year Campaign

 

 

 

 

 

 

Power Growth A Power Growth C
(Power at 50 years = 370 GWe) (Power at 50 years = 170 GWe)
ST U40g ST Us0g

Cumulative A Committed Cumulative Committed

Reactor Type Requirement Requirement* Requirement Reguirement

PWR, No Recycle 2.04 x 10 4.08 x 108 1.02 x 106 1.86 x 108

PWR, Uranium Recycle 1.63 x 10° 3.22 x 106 0.813 x 10% 1.47 x 10°

PWR, Pu Recycle 1.32 x 106 2.57 x 10° 0.658 x 10% 1.18 x 10°

CANDU, No Recycle 1.54 x 10° 3.17 x 10° 0.776 x 10° 1.44 x 10°

CANDU, Pu Recycle 0.939 x 105 1.85 x 10° 0.469 x 10° 0.846 x 10°
CANDU, Th-Pu Fueled

CR = 0.85 0.807 x 106 1.136 x 10° 0.392 x 10° 0.519 x 10°

CR = 0.93 0.549 x 10° 0.695 x 10° 0.262 x 10° 0.318 x 10°

CR = 1.0 0.503 x 10% 0.503 x 10° 0.230 x 10° 0.230 x 10°

 

*For 50-year reactor lifetime at 0.8 load factor.

LT-A
D-18

L
For the thorium-fueled CANDUs, much more of the total commitment would
be required at the end of the 50-year period. In this analysis, for
cases where the conversion ratio equals 1.0, the cumulative requirement

and the commitment are equal,

Fuel Cycle Cost Characteristics of CANDU Reactors

 

The fuel cycle costs for the CANDU reactor concepts are estimated using
a simplified fuel cost model discussed in Annex D.1l. Much of the input
required by the model is derivable from the data of Tables D.1 and D.3

of the fuel performance and fuel utilization sections of this Appendix.
Cost assumptions for reprocessing, fabrication, shipping and waste

disposal charges are from Appendixes H and I of this report.

These cost and fuel material flow requirements for the equilibrium

refueling cycle are shown in Table D.8 for several CANDU fueling options

and for comparison, those of a typical PWR reactor. It is apparent from .
the table that a substantial cost penalty will occur in the fabrication .
of either uranium~plutonium, thorium-U-233, or thorium-U-233-plutonium
fuels. The fabrication cost ratios of these different fuels may be of
greater interest than the estimated magnitudes shown in Table D.8.

These ratios are listed in Table D.9.

The estimated reprocessing costs for the PWR were based on the AGNS
plant expanded to handle the conversion of uranyl nitrate to UFg,
plutonium nitrate to PuO,, and the solidification or containment of all
radioactive liquid, gaseous and solid wastes from the reprocessing plant
operation. They are based on 5 MT/day plant specifically designed to
handle a particular fuel type. As such within the limit of accuracy of
the estimation process, no significant cost differential was found for
any of the several fuel types that might be considered by either the
CANDU or the PWR. As a result the basic reprocessing cost was taken to
be $226/kg. Since the recovery of the fissile content of thorium fuels

that are clad with zirconium is poorly understood, fuel cycle cost

 
 

 

Table D,8. Data for Equilibrium Fuel Cycle Cost Calculations of CANDU Reactors? !
Annual U30s Annual ThO; Separative Sum of Fab., Topping In-Core Shipping Plus Permanent
Feed Rate, Feed Rate, Work Req'd., Fabrication Shipping and Fissile Plus Reproc. Plus No Fuel Assembly
ST U304/ ST ThOz/ MT SW/ Rate,** Reproc. Costs, Pu Feed, Ex-Core*** Pabrication Waste Disposal Storage Costs,
Case GWe-Year GWe-Year GWe-Year MT/GWe-Year $/kg U g/kg HM  Time, Year Cost, $/kg Costs, $/kg $/kg
CANDU, No 168 0 0 127.7 132.5% 0 2,918 79.5 - 50
Recycle
CANDU, Pu 94 0 0 54.1 5388 0 4.29 312 226 -
Recyclett
A 0 107 o 97.3 611 o} 3.12 385 226 -
B 24,7 53.6 0 48.7 611 1 4.32 385 226 ‘ -
c 32.2 42.8 0 38.9 611 2 4.92 385 226 -
D 44.9 32.5 0 29.5 3Nl 4 5.82 385 226 -
E 53.4 26.7 0 24.3 611 6 6.62 385 226 -
F 59.9 24.3 0 22,1 611 8 7.12 385 226 -
1 31.2 71.4 31.2 64.9 611 0 3.44 385 226 -
2 40.2 39.6 40.2 36.0 611 o 9.66 385 226 -
3 52.6 24.3 52.6 22.1 611 0 6.39 385 226 -
4 0 126.0 0 114.5 611 0 3.21 i85 226 -
5 21.9 39.6 21.9 36.0 6ll 0 6.03 385 226 -
PWR, No 210.57 0 131.2 27.65 250t 0 4.82 150 - 100
Recycle
PWR, Uranium 162.84 o 131.2 27.65 376 0 4,82 15¢ 226 -
Recycle
PWR, Pu 129,18 0 96,2 27.65 469,588 0 4.82 500 MO, +
Recycle 150 VO3

AV. = 243.5§§

 

*All data refer to an 0.8 reactor load factor.
**This is also used for the reprocessing and shipping rate.
***Ex-core time = 700 days.
tIncludes $100/kg perpetual storage costs for PWR and $50/kg for CANDU.
titCases A to F represent the CANDU-Th Pu makeup concepts, and Cases 1 to 5 represent the CANDU-Th U-235 makeup concept described in Table D1,
SAssumes all of the reload batch is of mixed oxide.
§5Assumes only 0.267 of a reload batch is mixed oxide.

61-a
D-20

Table D.9.

Fabrication Cost Ratios of CANDU and PWR Fuels

 

PWR, slightly enriched UOj

PWR, slightly enriched UO;-Pu0O,

CANDU, natural uranium (UO5)

CANDU, slightly enriched (UO,-Pu0,)
CANDU, slightly enriched (UO,-PuO,-ThO,)

1.00
3.33
0.53
2.08
2.57

 

 
D-21

estimates were also made for a 50%Z increase in the reprocessing charge
when applied to thorium fuels. In addition to the unit costs listed in

Table D.8, the remaining basic cost assumptions of the study are:

Separative work $70/SwWU
ThO, $15/1b ThO,
Inventory charge rate 15%/vyr
Reactor load factor 0.8

Figure D.6 presents a comparison of the fuel cycle cost of the PWR and

the CANDU when both reactor types are not fueled with thorium.

This figure shows that for a large U30g cost range, the CANDU no recycle
concept shows potential for lowest fuel cycle costs. In these estimates
the cost of D,0 initial inventory and annual losses have not been included.
Other studies* have estimated the D,0 loss cost to be about 0.35 mills/kWhr
(for D,0 at $110/kg and 16%/yr charge rate) and the initial inventory

cost to be about 2.5 mills/kWhr. The dashed lines show the effect of
adding the annual D;0 loss cost to the CANDU fuel cycle cost. Even with
this cost added the CANDU, no recycle fuel cost is less than the PWR
plutonium recycle cost when the cost of U30g is less than $25/1b.

As presently conceived (and understood by this author) the Canadian
concept for plutonium recycle will never be economically competitive.
This fueling concept adds plutonium to all reload fuel assemblies to
increase the fissile loading to near 1% such that the achievable burnup
approaches 18-19 MWd/kg HM. As such, this concept pays a high plutonium

fabrication cost for all assemblies. It is possible that the use of

 

*
A Brief Survey of Considerations Involved im Introducing CANDU Reactors
into the U.S., Argonne National Lab., December 1975 (unpublished).
ORNL-DWG 76-17693

 

 

0 I I T
—==INCLUDES D,0 OPERATING LOSS
COST OF 0.35 mills/kWhr
e -
PWR, NO RECYCLE

 

 

' |
PWR, URANIUM RECYCLE
8 l -~\T\\§;

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

= CANDU, ,
$ PLUTONIUM RECYCLE
§
E 7 / 1' — /7,
% /)
8 /
“os 1 4
O > 7
5 -~ 7/ /
: Y
2 . /
( N~PWR, PLUTONIUM
P RECYCLE
// CANDU, NO RECYCLE
4 V
3
0 25 50 75 100

Uz0g COST (¥/1b)

Fig. D.6. Comparison of fuel cycle cost of the PWR and CANDU as
a function of U304 cost.

 
D-23

plutonium spiking and slightly enriched uranium assemblies (as contem-
plated by the PWR) might produce more favorable economics. On an economic
basis the PWR with plutonium recycle is very competitive with the CANDU.
Uranium recycle only in the PWR is seen to be economically justified at

a U30g cost of well under $20/1b.

Figure D.7 compares the fuel cycle costs (not including D,0 losses) of
concepts on uranium or thorium fuel options. 1In all circumstances
considered, the CANDU-Th concept having a conversion ratio equal to 1.0
was uneconomical. Based on Figure D.7 the preferred method for utiliza-
tion of thorium in CANDUs is to use highly enriched U-235 as the makeup
fuel. This concept (at CR = 0.85) was the lowest cost option studied

and was remarkably insensitive to U30g cost increases. The use of
plutonium as the makeup fuel was found to be considerably less economic.
However, these calculations were basgd on unit costs of Pu as recovered
from CANDU reactors. Use of Pu from LWRs (if available) or from slightly

enriched CANDU cycles would give more favorable results for Pu use.

Figure D.8 shows the fuel cycle cost as a function of the conversion

ratio for the CANDU thorium concept having plutonium makeup. Distinct
minima were found at about 0.9 to 0.92 conversion ratio. The effect of
increasing the reprocessing cost from $226/kg to $339/kg (a 50% increase)
is also shown. This increase tends to drive the optimum conversion

ratio down. Conversely, increasing the cost of U30g tends to increase

the optimum conversion ratio.

Figure D.9 presents similar results for a CANDU thorium concept having
937 U-235 as the makeup fuel. For this makeup fuel the optimum con-
version ratio is seen to be less than 0.8. This figure also shows the
effect of increasing the power density on both conversion ratio and fuel
costs. At the higher power density it seems apparent that conversion
ratio of 1.0 is probably not attainable. The lower power density concept
seems to have the potential for somewhat lower fuel cycle costs at its

optimum conversion ratio.
ORNL-DWG 76-17694
10

 

CANDU, Th-Pu,

CR=1.0-
9 \

 

 

 

 

 

CANDU, Th-23%y,
CR=09—_]

7 >/
CANDU, Th-Pu, /
CR=085

/

CANDU, Th-Pu
. CR=093

CANDU, Pu /

RECYCLE -
f

CANDU, NO
RECYCLE

CANDU, Th-2%%y,
CR=085

 

 

\\
\

FUEL CYCLE COST (mills/kWhr)

 

 

R

\

 

 

 

 

 

 

 

0 25 50 75 100
UzOg COST ($/1b)

Fig. D.7. Comparison of fuel cycle cost of CANDU concepts on
uranium or thorium fuel options.

 
D-25

ORNL-DWG 76-17695

]

FABRICATION = $385/kg

REPROCESSING = $226/kg

_ FABRICATION = #385/kg
REPROCESSING = #339/kg

{1

 

10

 

 

 

   

 

UsOg COST

$100/1b

 

FUEL CYCLE COST (mills/kWhr)
o

 

 

 

 

 

 

 

 

 

5
0.75 0.80 0.85 0.90 0.95 1.00

CONVERSION RATIO

Fig. D.8. Fuel cycle cost for CANDU~thorium concept having
plutonium topping.
D-26

ORNL-DWG 76-17696

 

 

 

 

 

 

   
  
  
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

12 '
FABRICATION = $385/kg
/' REPROCESSING = $226/kg
_ FABRICATION = '385/kg
= REPROCESSING = $339/kg
10 . |
T 25 6 Mith
£ MT
|
x 9
0
e
th 8
o
O
Y
o 7
>
O
-
u
S 6
5
4

 

 

 

 

 

 

 

 

 

 

3
070 075 080 085 090 085 100

CONVERSION RATIO

Fig. D.9. Fuel cycle cost for CANDU-thorium concept having 93%
U-235 topping.

 
D-27

ANNEX D.1

CALCULATIONAL MODEL FOR EQUILIBRIUM FUEL CYCLE COSTS

The cost model employed in this work is highly simplified, but does

consider the basic costs incurred by a utility. 1In this model no dollar

value is attached to recycled fissile material. Present value discounting

is not done, but an inventory charge based on total in-core plus ex-core

time is estimated.

The burnup cost (of either U30g or ThO;) is defined by the equation

B=ped . p_ST 2000 1b 10° mills _GW _ _yr D
Ib GWe-yr ST $ 10°5 kW 365 D 24 hr

2.28311 x 107" Fc « F, mills/kWhr .

The separative work cost is defined as

MT SW_ | $ . 10° mills _GW__ 103 kg yr D

5 =X " GHe—yr . T Kp-SwU $ 105 kW MI 365 D 24 hr

1.14155 x 1074 X « Y, mills/kWhr .

The fabrication, reprocessing, shipping or waste disposal costs are

defined as

 

c =M. $ . 103 mills GWw _ 103 kg yr D
X GWe-yr x kg HM $ 106 kW MT 365 D 24 hr

1.14155 x 1074 M - F_ mills/kWhr .

The cost of supplemental plutonium feed (used for topping in some
CANDU-Th concepts) is based solely on the money invested in recovery
of plutonium from a standard natural uranium CANDU that would not

otherwise be recovering the plutonium. Thus, it is defined as:
where

Vp

rec

where

Fr =

Fs =

. Cp

D-28

 

Mp BLEPU oo § ) MTHM 103 kg 103 mills = yr  _ D
kg HM gm GWe-yr MT $ 365 D 24 hr

1.14155 x 10~* Mp « Vp * M, mills/kwhr ,

r $ . kg HM
rec kg HM 2.7 gm fissile Pu

 

0.37037 F___, $/gm.

Net recovery cost of CANDU plutonium,

Fr + Fw - Fs, $/kg HM,

Reprocessing cost,
Waste and safeguards handling costs,

Cost of perpetual storage of spent CANDU fuel assemblies.

= 4,22796 x 107> Mp « M F o> @ills/KWhr .

E. The inventory cost is based on the average value of all other fuel

cycle

charges prorated over the entire in-core plus ex-core time

and is defined as:

=
fl

where

Cx

3
|

(B+S + Cx + Cp) % « I+ T, mills/kWhr ,

Sum of fabrication, reprocessing, shipping and waste disposal
costs,
Annual charge rate, 7%/yr,

Sum of equilibrium in-core plus ex-core fuel times, years.
D-29

ANNEX D.2

CANDU-PLUTONIUM RECYCLE DATA

Subsequent to the analysis made in this report and as a result of
further discussions by ANL with the Canadians, the plutonium recycle
concept for CANDUs has been updated.* The revised recycle concept
considers the self-generated recycle of plutonium. Some of the
pertinent fuel characteristics of the ;revious and present recycle

concept are listed in Table DA.1.

The uranium requirements for the updated recycle mode are 827 of

those estimated originally. Relative to the no recycle mode, this
method of plutonium recycle provides a 50% fuel saving as compared
with 60% for the recycle concept given in Table D.1l. This result

does not change the conclusions previously obtained.

 

*
Personal communication from Edward M. Bohn, ANL, to P. R. Kasten,
ORNL, September 22, 1976.
D~30

Table DA.1. Comparison of Pu Recycle in CANDU Reactors

 

Reactor System
in Table D.1.

Self-Generated
Recycle System

 

Specific Power, kWth/kg HM
Inventory, MT/GWe

Discharge Burnup, MWd/kg HM
Fuel Residence Time, Years

Equilibrium Cycle Loading,
MT HM/GWe-year
U0,

U-Pu-0»y

Fissile Enrichment, wt 7% HM
U-235 in UO»

Fissile (U + Pu) in U-Pu-0,

Equilibrium Cycle Discharge, wt 7
U-235 in U0y

Fissile Pu
U30g Requirement, ST?/GWe-year
30-year Commitment, ST/GWe

26

128

16
2.37

31.5
40.1

0.711

(1.26 + 0.71)

0.17
0.33
94

3100

26

128

18
2.1

59.8

0.711
1.02

0.11
0.35
79
2540

 

aShort tons

 
l-

D-31
DATA SOURCES

E. Critoph, et al., Prospects for Self Sufficient Equilibrium Thorium
Cycles in CANDU Reactors, Tran. ANS, November 1975.

J. S. Foster and E. Critoph, The Status of the Canadian Nuclear
Power Program and Possible Future Strategies (a paper prepared for
the Wingspread Conference of "Advanced Converters and Near Breeders,"

May 14-16, 1975).

S. R. Hatcher, Thorium Cycle in Heavy Water Moderated Pressure Tube
(CANDU) Reactors, Tran. ANS, November 1975.

S. Banerjee, E. Critoph, and R. G. Hart, Thorium as a Nuclear Fuel
for CANDU Reactors, Canadian J. of Chem. Eng., Vol. 53, pp. 291-296,
June 1975.

A. M. Perry, An Analysis of Heavy Water Reactors Operating on the
Thorium Cycle, Inst. For Energy Analysis (two unpublished papers
analyzing uranium requirements and fuel cycle costs of the CANDU-Th
with U-235 topping); personal communication to E. H. Gift, March 17,
1976.

C. L. Moon, Pickering Generating Station, Nuclear Eng. Int., pp. 501-
515, June 1970.

M. F. Duret, Plutonium Recycle in CANDU Type PHW Heavy Water Reactors,
AECL-3910, May 1971.

A. J. Mooradian and O. J. C. Runnalls, CANDU--Economic Alternative to

the Fast Breeders, AECL-4916, September 1974.

W. B. Lewis et al., Large Scale Nuclear Energy from the Thorium Cycle,

A/CONF. 49/P/157, Geneva, 1971.
10.

11.

12.

13.

14.

D-32

A. W. L. Segal, Estimating CANDU Fuel Costs, AECL-4273, September
1972,

B. I Spinrad, Thoriwm Based Fission Reactor Fuels, Conf. on Environ-
mental Aspects of Non-Conventional Energy Sources, Denver, Colorado,

March 1976.

D. D. Stewart, The Canadian Incentive for Fuel Reprocessing and
Plutonium Recycle, AECL-3136, June 1968.

A Brief Survey of Considerations Involved in Introducing CANDU
Reactors tnto the U.S. (draft), prepared by the staff of Argonne
Natl. Lab. for DNRA of ERDA, January 5, 1976 (report).

Study of Fission Power Reactor Development Strategy for the United
States (draft), in preparation by the staff of Battelle Columbus

Laboratories for the National Science Foundation.

 
APPENDIX E

THORIUM FUEL CYCLE IN FBRs

Summary: Several alternatives are available for use of thorium fuel
cycles in FBRs. Thorium can be utilized in the FBR core, blanket, or
both. Use of thorium in the core results in reduced breeding gain
compared with the uranium-plutonium fuel cycle, but lower specific
inventories might be possible if high power densities are practicél.
Under the latter circumstances, the doubling time of the two fuel cycles
would be-comparable. If core power densities are about the same value,
the uranium cycle would have lower doubling times than the thorium

cycle. Use of thorium in the core results in significantly more negative
sodium void coefficients of reactivity for LMFBR cores. The lack of
reprocessing experience on thorium containing FBR fuels, and the lack of
irradiation performance data on thorium metal fuels are primary impedi-
ments for use of thorium FBR fuels. Incentives are the improved reactivity
coefficients possible with thorium fuels, possible application of the
thorium cycle in '"denatured" fuel usé, and in fast and thermal reactor

fuel cycles once the fissile resource problem is solved.
Thorium Use Options in FBRs

Thorium can be utilized in an FBR in four modes: 1) 233y-Th fuel can be
used in the core with thorium blankets so that no plutonium or natural
uranium is involved in the breeder fuel cycle. 2) 239pyu-233y_Th fuel
can be used in the core with variations on relative 23%Pu to 233y

ratios and blankets can contain thorium or uranium. 3) A thorium radial
blanket can be used with a 239Pu-238y core that has axial blankets of
uranium as proposed for the GCFR;1 4) A thorium radial blanket and a
2337-238y core. Each of these systems has its own advantages and
disadvantages, many of which are described in this Appendix in comparison
with the conventional LMFBR or GCFR designs that incorporate 239py-238y
cores with uranium blankets. Additional information on "denatured" fuel

use is given in Appendix Q.

E-1
E-2

Performance and Safety Considerations

 

In Appendix A it was shown that for identical fuel forms and coolants
the uranium-plutonium cycle would have the largest breeding gain. It
was also shown that thorium metal fuels have better breeding performance
than thorium oxide fuels. A key point for thorium utilization in FBRs
is that thorium can be used as a metal fuel whereas the irradition
properties and temperatures of phase change of uranium metal severely
restrict its usefulness. Some selected properties of thorium and
uranium fuels are shown in Table E.l1. The phase change temperature and
melting temperature of Th metal are significantly higher than those of

U metal. In addition the Th matrix can contain significant fractions of
U and Pu without significantly affecting its melting point. Uranium
metal has anisotropic properties such that severe swelling and distortion
take place under thermal cyclifig and reactor irradiation. Uranium
alloys that have been developed to reduce irradiation distortion still
show fuel swelling of about 10 vol 7 per atom percent burnup. Thorium
metal with up to 20% uranium has shown excellent radiation stability in
thermal neutron irradiation with volume increases of 2.5 vol % per atom
percent burnup at 650°C, for burn-ups up to 4 atom percent. Thus thorium
metal and thorium-Pu-U alloys show promise for use in FBRs. Irradiation
experience with thorium metal in FBR enviromments is, however, very
limited. The use of thorium metal as fuel in FBRs also offers the
possibility of using cladding alloys other than stainless steel. A
vanadium-207% titanium alloy has been suggested as a possible candidate
because of its compatibility with thorium metal and its superior irradi-

ation behavior.?

The use of this alloy in a sodium envionment would
require that the oxygen content be kept very low. The neutron absorption
of this alloy is less than stainless steel so that improved breeding is
feasible. There is, of course considerable uncertainty in the practi-
cality of thorium metal use, and that should be borne in mind in the

following discussion,

The performance of FBRs fueled with mixed oxides of Pu and U, mixed

oxides of 233y and Th, and Th metal fuels has been recently studied.?

 
Table E.1. Selected physical properties

of thorium and uranium fuels

 

 

UG, ThG, U(mctal) Th{metal) uC UG, ThC ThC;
Melting point (*C) 2750 3290 1130 1700 2320 2480 2625 2655
Melting point (*F) 4980 5970 20170 3100 4200 4600 4760 4510
Deasity (room temperatuie} {(g/em?) 10.5 9.7 19.0 1.6 13.0 1n 10.6 9.6
Thermatl cm‘:duc!ivirya =
at 650°C (W/em °C) 0. 035 0.0i0 10,37 0.45 0.23 ~0.2 ~0.25 I
3t 650°C (Buu/hr f °F) 2.0 2.3 21 26 13 ~12 ~ 14 w
Tempciature at which phase 665 (a 10 B) 1375

change occurs ("C) 75 (Fto )

(FCC to BCC)

 

# Ceramics generally suffer a decrcase in conductivity with long tcactor exposute at relatively low temuperature, which is not considered in the above values,

At high temperatutes (> ~ 1700°C), irradiation effects on k do not appear significant,

SOURCE: P. R. Kasten, '"The Role of Thorium in Power Reactor Development,' Atomic Energy

Review, vol VIII, No. 3, p. 473.
E-4

The neutronics calculations were performed for spherical reactors with

37 V% fuel, 17 V% structure, and 46 V7 coolant in the core region. Because
the calculations assumed spherical cores, the breeding ratios are higher
than in practical systems and the calculated fissile masses are.lower

than in practice; however, because all the studies were performed with
consistent bases for core power densities, relative performance comparisons

can be made. Table E.2 lists the results of this study.

Cases No. 1, 4, 7, and 11 show the potential capability of 233U-Th
breeders with metal or oxide fuels and Na or He coolants. These systems
do not involve any use of Pu or natural U. As a basis for comparison
case No. 10 is the reference LMFBR with Pu-U oxide core and U blankets.
The reactivity effect of complete sodium voiding for the 233y-Th systems
is strongly negative as compared to the approximately $11 of positive
reactivity that is generated in the reference LMFBR. The breeding ratio
of the 233y-Th metal, Na cooled core (case 1) is .09 less than the
reference LMFBR (case 10); however, the fissile mass of the 233y-Th

case is only 70% of the fissile Pu mass in the reference case. Thus the
doubling times of these two systems would be nearly identical. There
are no current detailed evaluations of the technical problems associated
with Th metal fuel development for FBR. These preliminary neutronics
studies suggest that such evaluations should be undertaken to assess the
feasibility, quantify the benefits, and determine the necessary develop-
ment of Th fuel for FBRs. 1In particular, it is not certain that the core
power densities assumed in these preliminary studies are practical for

the metal fuel cases. The values are prébably too high.

A second class of Th utilization in FBR involves the use of Pu-U-Th
fuel alloys. Cases 2, 3, 5, and 6 from Table E.2 demonstrate the per-
formance of these metal fuels. Again the inclusion of Th leads to a
negative sodium void coefficient for total core voiding. The 239Pu-
2337-Th system with a U metal blanket (case 3) shows a nearly self-
sufficient Pu core and a large net 233y production. As a matter of
comparison a 1000 MW(e) high gain (CR = 0.84) HTGR requires only 157
kg/yr of 233U fuel. Thus the Th based LMFBR could supply the fissile

 
8

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

/
Table E.2. Characteristics of thorium and uranium fast breeder reactors
Extra
&k for Fertile*™ Flss{le Materfal
) Core Core Avg Coolant_| Captures Produced {(kg/yr)
Case Cladding[ 8lanket Core Fissile | Volume {Power Density | Fissfile Mass | boppler | Votding® Fissile 213 219
Mo, | Core Fuel Materfal | Material| Paterial | Coolant Hateria) (liter) J[MW@h)/liter) {kg) Coeff {3) Absorptions v Pu
1 | Th Meta) Alloy ¥-20 T4 | Th Fetal Ka 2y 4003 0.62t* 1572 -0.009 -4.60 1.26 208.7 -
2 . - . . 13p23yytt - 0.62 1343 Py | -0.00% | -2.2% 1.38 B97.2 | -598.7
+425 133y
) e . Ukel | . . 0.60 1383 Pu_ 1-0.009 | -0.71 1.48 482.0 | -4%?
+435 13y
4 . . ™ Metal | He 33y 5456 0.45 2143 -0.007 | -0.13 1.29 2241 -
s - . . - pp2iYY . 0.45 18 Py | -0.007 | «0.01 1.45 $M.6 | -584.3
+570 2%y '
6 . . U Petal . - . 0.42 1831 Py | -0.010 | +0.02 1.55 486.5 -44,2
485 133y
7 P Th WX $S ™0, Ka 2330, 631 0.39"t 1879 -0.016 -2.17 1.16 130.7 -
8 . . . . 120, . 0.39 2466 -0.014 | +0.50 1.18 1103.2 | -959.6
9 - . o, . . . 0.38 2457 -0.014 | +0.70 1.20 769.7 | -589.7
10 U mx . - . - - 0.38 2223 Py - +31.73 1.35 - - 358.2
+129 2%y '
1N | Th MOX 5 Tho, He M0, 10,306 0.24"t 2910 -0.011 +0.01 1.21 169.9 -
12 . - . . 13%p0, . 0.24 m -0.010 | +0.26 1.27 1149.7 | -925.5
13 . . w, * - . 0.24 3593 -0.010 +0.27 1.3 821.6 ~564.9
4 jU WX . . . . - 0.24 3178 Py - +0.39 1.44 - 419.5
4203 HSu

 

 

-
N2 is voided completely from the core region and He is voided conpletely from all regions.

JTHOX = Mixed Dxide,
This is the norwal breeding ratio for the systea in which the fissile material produced 13 the same as that consumed, . The values shown are for a
spherical reactor and would be smaller for the uwsual cylindrical resctor.

he ™-U-Pu 8110y contains 9.5 wI 33%py.

?hese power densities are the same 3s in the deronstration-iize desfgns of the LMFBR and the 6CFR,
1 Inis value of the power density 45 less than that for the ®arly EBR-II U-metalefueled cores.

SOURCE :

B. R. Sehgal, C. Lin, J. Noser, W. B. Loewenstein, "Thorium-

Based Fuels in Fast Breeder Reactors," Trans. Am. Nucl. Soc.

21: 422 (1975).

 
E-6

material for itself and for two to three HTGRs. This high breeding

performance was obtained with a Th loaded FBR core that had an overall
negative sodium void coefficient. It should be emphasized that total
sodium voiding is usually not the worst case from the viewpoint of re-
activity changes. Further studies of cylindrical cores by the authors
of the above comparison study showed that for case 3, the "worst case"
reactivity increase was only 1/3 of that in a Pu-U LMFBR.3 For case 1

the "worst case'" reactivity increase was 1/12 of that in a Pu-U LMFBR.

The third use for thorium in FBRs is as a radial blanket material for

233y production. In the case of mixed oxide FBRs the substitution of a
thorium blanket for a uranium blanket does not significantly affect the
overall breeding performance, although there is a small decrease. Thus the
the selection of uranium or thorium blankets for an LMFBR depends on the
relative economics of the bred materials and on the availability of the

necessary reprocessing and refabrication technology.

Another possible FBR fuel combination involves use of U-233 and U-238

in the core, and thorium in the blanket. This corresponds to a mixed
cycle with Pu generated in the core. This type of fueling might be
utilized if '"denatured'" fuel containing less than 20% fissile uranium
were imposed on specific FBRs. The nuclear performance would be somewhat
better than the FBR fueled with U-233 and Th in the core and Th in the

blanket. More information on this is given in Appendix Q.

 
REFERENCES FOR APPENDIX E

R. J. Cerbone, N. Tsaulfanidis, '"Thorium Utilization in Gas-Cooled

Fast Breeder Reactors," Trans. Am. Nucl. Soc. 22: 703 (1975).

B. R. Sehgal, C. Lin, J. Naser, W. B. Loewenstein ''Thorium-Based
Fuels in Fast Breeder Reactors,' Trans. Am. Nucl. Soc. 21: 422

(1975).

B. R. Sehgal, J. Naser, C. Lin, W. B. Loewenstein, "Thorium
Utilization in Fast Breeder Reactors,'" Trans. Am. Nucl. Soc.

22: 704 (1975).
APPENDIX F
USE OF 233y AND 238y IN FAST BREEDER REACTORS
("Denatured" Fuel Cycles)

I. Introduction

The purpose of this appendix is to examine the feasibility of using
23333-238y pxide fuel in fast breeder reactors. By limiting the 233U
content to less than 207 of the uranium, the safeguard requirements
should be reduced significantly, since such a fuel system would avoid

the possibility of chemical separation of fissile isotopes of the initial
fuel, and the enrichment limitation is considered not to lend itself for
use in a nuclear weapons device. The Pu bred in these reactors is
assumed to be separated from the spent fuel and utilized in FBRs located
inside safeguarded areas. The investigations here centered around two
commercial-sized [1200 MW(e)] conceptual breeder designs; an LMFBR core
design from GE, and a GCFR core design from GA. :In both breeders,
calculations were first performed for the Pu-238U design and then for
233(j-238y fyel for comparative purposes. This appendix gives preliminary

results of the comparison.

II. LMFBR System

The LMFBR model used in this investigation is a GE design selected for
benchmark use by the Large Core Code Evaluation Working Group. It
consists of a two-enrichment zone core of PuO, and depleted 238U0,. The
core composition is for the beginning of an equilibrium cycle. A more
complete design description is given in Table F.l. Following a two-
dimensional diffusion theory (PO transport - corrected) calculation for
the Pu-238y core, 233y-238y fyel was substituted for the Pu-238y case.
The average fissile core enrichment dropped by 1.5%Z while the breeding
ratio dropped by 12% as a result of changing from Pu-238U fuel to a
233y-238y fyel. Detailed results are shown in Table F.2. The Na in the
beginning-of-life composition inner core zone was removed in both fuel

systems in order to estimate the effect of 233y on the Na worth. Although

F-1
Table F.1.

LMFBR Design Parameters

 

 

Reactor Power, MW(t)

Reactor Power, MW(e)

Core Height, cm

Core Diameter, cm

Core volume, &

Radial Blanket thickness, cm

Axial Blanket, thickness, cm
(each end)

Composition
Core
Axial Blanket
Radial Blanket

Enrichment (% fossile in total

heavy metal)
Inner Core
Outer Core
Core Average
Pu Composition
7 239%y
o 240Pu
7 21+1Pu
g 242p,
Inventory (tonne)
Fissile

Fertile

Volume Fractions

Fuel
Sodium

Structural

3085
1200
107.72
286.30
6940
38.91

33.35

PUO, + depleted 238y0,

Depleted 238UO;;_
Depleted 238U02

10% (239%pu + 241py)

13.4% (%3%u + 241py)
11.6% (23%u + 242py)

607%
23%

3.07

67.39

Core and Axial
Blanket
0.306
0.365
0.329

Radial Blanket

0.481
0.274
0.245

 

 
Table F.2. IMFBR Calculational Parameters

 

233

 

 

Fuel Pu (Reference) Tu

keff 1.037 1.037
Enrichment
(% fissile by atom)

Inner Core 10.0 8.7

Outer Core 13.4 11.6

Core Average 11.6 10.1
Breeding Ratio*

Inner Core 0.549 0.525

Outer Core 0.441 0.300

Total Core 0.990 0.825
Axial Blanket 0.163 0.190
Radial Blanket 0.222 0.198

TOTAL 1.375 1.213
*238U breeding 9Pu + 240Pu breeding 241Pu.
F-4
®

both cases, the increase for the 233y-fueled system was an order of .

voiding the inner core increased the effective system multiplication in

magnitude less than for the Pu-fueled system. A very preliminary burnup
study for the 233U-fueled case indicated that the 233U/23%u ratio in
the core approached 1/1 over the equilibrium cycle. Thus, while the
void coefficient of reactivity would increase over that of the initial-
fueled core, the reactivity change associated with sodium voiding in an
equilibrium core would still be only about half that of the plutonium-

fueled system.

 

IIT. GCFR System

The GCFR model used in this investigation was contributed by GA. It

consists of a four-enrichment zone core of PuO, and depleted 23800,

surrounded by an axial blanket of depleted 23800, and a radial blanket

of ThO,. The core compositions used were for the initial loading. A

more complete model description is given in Table F.3. Similar to the . .
LMFBR case, a two-dimensional diffusion calculation (Po transport-

corrected) was performed for the Pu-238U fuel system and than a criti-

cality search was performed for the equivalent 233y-238y gystem. The -
average percent fissile core enrichment dropped by 1.8% and the breeding

ratio dropped by 13%, as a result of changing from Pu-238y to 233y-238y

fuel. Details are given in Table F.4. When an equal-fissile inventory

233y-238y case was run (atoms 233U = atoms 23%u + 24lpu, and 238U was

added to return to the original k the breeding ratio was only 8%

eff)’
below the original Pu-2387 case.

IV. Conclusions and Observations

 

The initial calculations for 233y-238y fueled breeders indicate that a

breeder with acceptable breeding gain can be designed. In these initial

calculations the only design change was the fuel substitution and en-

richment change to achieve constant k effective. A proper evaluation

must include optimization of each of the candidate designs within the .,

same constraints and performance goals. The present results indicate
Table F.3.

GCFR Design Parameters

 

 

Reactor Power, MW(t)

Reactor Power, MW(e)

Core Height, cm

Core Diameter, cm

Core Volume, £

Radial Blanket thickness, cm

Axial Blanket, thickness, cm
(each end)

Composition
Core
Axial Blanket
Radial Blanket

Enrichment (% fissile in total
heavy metal)

Core 1

Core 2

Core 3

Core 4

Average Core
Pu Composition

g 239y,

7 240Pu

o 241Pu

o 2'42Pu
Inventory (tonne)

Fissile

Fertile

Volume Fractions

Helium
Fuel

Structural

3158
1200
135.2
321.7
10,980
34.53

33.35

PuO, + depleted 438y0,
PuOy, + depleted 238U02
ThO,

12.3% (23%u + 2%lpy)
14.0% (23%u + 24lpy)
16.6% (23%pu + 24lpy)
18.4% (239%u + 24lpy)
14.8% (23%pu + 2%lpy)

677%

267%

5%

2%

3.23

75.53

Core and Axial
Blanket Radial Blanket
0.640 0.34
0.221 0.50
0.139 0.16

 
F-6

 

 

 

Table F.4. GCFR Calculational Parameters
Fuel Pu (Reference) EEEQ
keff 1.024 1.024
Enrichment
(%4 fissile by atom)
Core Zone 1 12.3 10.5
Core Zone 2 14.0 12.0
Core Zone 3 16.6 14.3
Core Zone 4 18.4 15.9
Core Average 14.8 12.7
Breeding Ratio*
Core Zone 1 0.328 0.296
Core Zone 2 0.206 0.180
Core Zone 3 0.100 0.086
Core Zone 4 0.091 0.077
Core Total 0.725 0.639
Axial Blanket 0.388 0.327
Radial Blanket 0.307 0.268
TOTAL 1.421 1.234

 

*238

U breeding 239

Pu, except in radial blanket, where 232

Th breeding

233

U.

 
F-7

that such designs would be practical for both LMFBR and GCFR breeder

reactors.

Assorted observations from the initial study are as follows:

1.

233y provides a lower critical mass than Pu because of a
higher fission cross section, but depresses the breeding
relative to Pu because of the lower eta values in the
important regions of the neutron energy spectrum.

The LMFBR and GCFR designs used here were those optimized
for Pu fuels. While substituting 233y fuel results in
reasonable breeding performances, the designs probably

are not optimal for 233y yse.

Current results indicate that variations in relative core
zone enrichments and in fuel density will improve nuclear
performance.

The use of 233U fuel in the LMFBR should reduce the sodium
void coefficient. Current results indicate that the
sodium void coefficient would drop approximately by factors
of 2 to 4 for the equilibrium cycle. The higher 238y fuel
content should lead to a slight improvement in the Doppler
coefficient.

The effects of fission products and control poisons which
were omitted in the initial calculations will not change
the overall conclusions. This is largely due to the high
conversion ratios for these large breeder cases.

The combined effects of 233U-238y fuel in possibly reducing
safeguards risks and in reducing sodium void coefficients
suggest that this fuel should be given further consider-

ation in FBRs.
 

 

 

T,

 

 

 

L e

 

 
APPENDIX G

ORE AND SEPARATIVE WORK REQUIREMENTS
IN AN INTEGRATED NUCLEAR ECONOMY

Summary: In several places in this paper ore utilization capabilities
of a given reactor system are described by calculating the ore require-
ments to provide the initial fissile inventory, and makeup inventory for
30 years of reactor operation. This information is useful, but does not
provide insight into the time variation of ore requirements. For example,
high gain converters have a high initial inventory compared to low gain
converters, and ore requirements in the early years of operation are
therefore higher for the high gain converters. Another shortcoming of
the "reactor commitment'" method of describing ore utilization is that it
does not permit easy comparison of different strategies or reactor
mixes. To overcome this shortcoming a simple model has been developed
to evaluate the time variation in ore and separative work requirements.
This model is described in this appendix, and several example cases
given. The model for computation of cumulative ore requirements is
given by Eq. (Gl).

t

sp- [ 2Lac, (61)

Fm s D

where
Fm = cumulative ore requirements (kg) for a given reactor type,
S = specific inventory [kg ore/MW(e)],
P = installed electrical capacity [MW(e)] for a given reactor type,
D = doubling time for the given reactor type (years). This is a
negative number for converters.

t = years since initial installation of the reactor type.

If a linear power growth rate is assumed, the integral in Eq. (Gl) is

easily evaluated.

The model for computation of separative work requirements is given in

Eq. (G2):
kg Swu

where

V(X)

kg Swu

G-2

7 EY
V) - V() + ( V) - v, (62)

Xp = X3/

X

(2X - 1) lnm,

"value function" which represents the value of one unit
of uranium at enrichment X,

product enrichment (decimal),

tails enrichment,

feed enrichment,

kg of separative work per MW(e).

Five sample cases are given, which compare the ore and separative work

requirements of LWRs, HTGRs, and FBRs. The point is made that if FBRs

are delayed to the year 2000 or beyond, some form of high gain converter

is needed to permit nuclear energy generation at that time to continue

at the same level.

Model for Ore Requirement Computation

A simple accurate model of fuel resource requirements would be bene-

ficial to the understanding of the factors which influence fuel resource

requirements as well as for initial survey estimates. Fuel resource

requirements are currently estimated with complex computer programs such

as ALPS,1 which was not available to us in the time available for this

study.

A model of the mined fuel requirements is developed for a system of

similar reactors (LWRs, BWRs, or HTGRs, etc.) after which it is gener-

alized to include a mixture of reactor types. The greatest unknown

input to this model which has the most significant influence on fuel

resource requirements is the estimated total nuclear power production

 
G-3

growth and how it is divided among the reactor types. Several examples
will be given after the model development has been completed to illus-

trate this effect.

A fuel utilization model of a system of similar reactors can be developed
from a simple mass flow balance. Let time be divided into equal intervals
of length At. For the ith time interval, the reactor fuel inventory
required to produce a specified power is then:

I, =F +R, , +1, , (1-fAt),
where

fAt = fraction of fuel replaced each At,
1 - fAt = fraction of reactor fuel remaining each At,
f = fraction of reactor fuel replaced yearly,

At = incremental time stop (years),

F, = mined fuel for ith time interval,
; = reactor fuel inventory for ith time interval,
R/ x = fuel recycled from kth interval for use in ith interval,
but
I, =58P, ,
i
and

AL,

R, = (fAtIi_k +-——f5———> 1-8,

P, = power at ith time interval (power capacity),
S = fuel inventory per unit power,
B = fraction of fuel lost during recycle (assume B = 0),

fuel cycle inventory
rate of excess fuel production °

 
G-4
The total mined fuel requirements for the first m time intervals is just

the summation of Fi from i =1 to i = m. If this summation is carried -

out and At allowed to approach zero, the result is:

t 0
F(t) = § [P(t) - P(O) + £ j. P(x) dx - £ j- P(x) dx
t-t' -t'

t-t
1
- DI ' P(x) dXJ s
-t
where
F(t) = mined fuel requirements at time t (total mined fuel put

into reactors from time 0 to time t),
t' = recycle time (storage time + reprocessing time +

refabrication time).
If t' =0,

t
F(t) = SB(t) - SP(0) - = f0 P(x) dx .

Everything has been defined in the above equations except D. The constant
D depends on reactor type and can be defined on the basis of certain

characteristic reactor parameters.

- fuel cycle inventory
rate of excess fuel production

SP(t) x 1/1f x P(t)
(CR - 1) £Xcess8 kg % (1 + o) kg burned " kg fissioned " MW < 365 days
kg burned kg fissioned 1000 MWDt nMWT year

 

 

2.74 nS

D=t DO+ oD °

 
 

G-5
where
CR = f?ss%le produced (converstion ratio),
fissile consumed
n = thermal efficiency,
) 1f = load factor,
o = captures/fissions (in fissile isotope).

Therefore, to estimate the mined fuel requirements, values for the

following parameters are required:
kg fissile -
S( Mile > f(yr 1), t'(Yr), Cr’ n, a, 1f:

and

P(t) (MWe) for each reactor type.

The above derivation was for a system of reactors that were all of the
. same type. To determine the total mined fuel requirements in a nuclear
economy comprised of several reactor types, the fuel requirement for

each reactor type must be summed, or:

N
Foop(t) = i§1 Fj(t) ,

where
N = number of reactor types,

Fj(t)

mined fuel requirement for reactor type j in kg
of fissile fuel,

Ftot(t) = total mined fuel requirement in kg of fissile fuel.
G-6

Model for Separative Work Requirement Computation

The separative work requirement can readily be calculated from the
preceding model for the mined fuel requirement. The preceding model
calculated the mined fuel requirements in terms of fissile fuel placed
in each reactor type. Since separative work is a measure of the value
of fuel of a certain enrichment, the separative work requirement for
each reactor type can be determined. The total separative work require-
ment would then just be the sum of the separative work for each reactor

type.

Separative work is calculated as follows:?

 

Let
xp = product enrichment (decimal),
X, = tails enrichment,
x, = feed enrichment.
Then
X = X
D) F=°P _ILtfii_ ,
r T
where
F = flow weight of feed,
P = flow weight of product.
X
(2) V(x) = (2x - 1) &n T % °
where V(x) = "value function" which represents value of one unit of

uranium of assay x, and

 
G-7

X - X
- Xy _ u(X P__t -
(3) Swu P{V( p) - V(*t) + ( ~ ) [V (xp) V(xF)]} s
T %
SR = §%2~ [SR is a constant which depends on reactor type

(depends on fuel, tails, and feed enrichments
but feed and tails enrichments are assumed the

same for all reactors in the system).]

The total separative work units in an integrated nuclear economy is

then:

Swu(t) =

N
L Fj(t)SRj/xp ,
j=1 ]

number of reactor types,
mined fuel requirement for the jth reactor type in kg
of fissile fuel,

enrichment of fissile fuel,

defined above and for jth reactor type.

The total natural uranium requirement is:

FNat(t)

 

N Fj(t) (xpj - x'I‘)
= E - .
j=1 xp, *r T *r

J

The total U30g ore requirement is then:

Fore(t)

= 1.18 FNat(t).
G-8

Example Cases

Five example cases are presented. Each example is performed for various
total nuclear power rates from 1980 on. The amount of mined ore require-
ment is presented for each example and the separative work requirement

is calculated for several examples. Five reactor types are considered;
standard and high-gain HTGRs, LWRs (U and Th cycle), and FBRs. The
following values for the necessary reactor parameters were selected to

represent the reactor types.

Table G.1 Reactor Parameters

 

LWR LWR-Th HTGR HTGR FBR
(standard) (high-gain)

 

 

3 kg fissile in core
1.9

Mwe 2.85 1.39 2.23 2.56
Total fissile

Core fissile 1.5 1.5 1.34 1.45 1.5
CR .60 .70 .66 .82 1.25
n .33 .33 .39 .39 .39
a .27 .15 .15 .15 .25
D (yr) -11.91 -16.02 ~7.26 —23.82  +22.52%
Xp .003 .003 .003 .003

Xp .00711  .00711  .00711 00711

Xp .030 .9315 .9315 .9315

 

*
The FBR has a fast fission effect which decreases the parameter D by
about 207%. About 20% of the fissions are nonfissile fissions.

In these examples the ore and separative work values included in the
tabulations reflect commitment only to the year specified. The total 30
year commitment of ore and separative work for a given reactor are not

included.

 
G-9

Example 1

Example 1 models an all LWR economy from 1980 to 2030. Plutonium is

recycled on a two-year recycling time. The nuclear power capacity as a

function of time is shown in the figure.

Peo

Year

2000
2010
2030

Year

2000
2010
2030

 

LWR

 

19680

2030

Py = at -
A = 20,000, 3Q000, gd 40,000 MWe/Senr

i

Total U30g Requirement (Metric Tons x 10°)

a = 20,000 a = 30,000 a = 40,000
. 887 1.33 1.78
1.63 2.45 3.27
3.73 6.16 7.46

Total Separative Work (kg x 10°)

a = 20,000 a = 30,000 a = 40,000
.388 .588 0.787
.720 1.08 1.45

1.65 2.72 3.30
G-10

Example 2

In example 2, the standard HTGR, the high-gain HGTR, and the LWR are
compared on the basis of ore and separative work requirements for an
assumed number power capacity curve from 1980 to 2030. Plutonium and
U-233 are recycled on a 2-year recycling time. The power capacity

versus time is shown in the figure.

Py Pw= Po

 

 

 

c1==AK10CE>hAVME/AyeflY

Total U303 Requirement (Metric Tons X 108)

Year LWR LWR-Th HTGR (standard) HTGR (high-gain)

2000 1.78 1.59 1.01 1.05
2010 2.59 2.20 1.60 1.36
2030 2.43 1.83 1.76 .931

Total Separative Work Required (kg X 10%)

Year LWR LWR-Th HTGR (standard) HTGR (high-gain)

2000 . 787 1.19 .755 .785

2010 1.14 1.65 1.20 1.07

2030 1.07 1.37 1.32 .696

NOTE: The mined ore and separation work requirement is proportional .‘

to the power growth rate a.

 
G-11

Example 3

In example 3, the standard HTGR, the high-gain HIGR, and the LWR are
compared with a different assumed power growth. Again the mined ore and

separative work requirment is proportional to the value of a.

F%f) Pw) = Po

Q@,}

1980 2000

@ = 40,000 MWe/ year

U30g Requirement (metric tons X 106)

Year LWR HTGR (standard) HTGR (high-gain)
2000 1.78 1.01 | 1.05

2010 2.59 1.60 1.36

2020 3.40 2.18 1.67 -

2030 4,21 2.77 1.98

Separative Work Required (kg x 107)

Year LWR . HTGR (standard) HTGR (high-gain)
2000 . 787 . 755 .785
2010 1.14 - 1.20 1.02
2020 1.50 1.63 1.25

2030 ©1.86 2.07 1.48
G-12

Example 4

In example 4, the effect on mined ore required of the introduction of

the FBR and the phase out of the LWR is illustrated. The power capacity

versus time is shown in the figure.

P

 

 

Year
2000
2010
2030
2040
2050

2030

a = 40,000 MWe /vear

Total Uz0g Requirement (Metric tons x 10©)

LWR*
1.78
2.90
3.27
3.27
3.27

 

FBR**
0.0
(.308)
(.838)
(.266)
(-.482)

Total
1.78
2.90
3.27
3.27
3.27

*
The LWR is considered to recycle all Pu not used by the FBR.

* %

The numbers in parenthesis represent the equivalent U30g requirements
if U235 had the same nuclear properties as Pu in FBRs. Since FBRs
use Pu generated from LWRs, no mined U30g is needed for FBR fissile
requirements; however, the numbers given in parentheses do provide
perspective relative to fissile fuel use.

..

 
G-13

Example 5

In example 5, the effect of the introduction of the FBR in the year 2000 on
mined ore requirement is illustrated. ‘In this case, the LWR does not phase
out after FBR introduction but maintains a constant power capacity level.

The nuclear power capacity versus time is shown in the figure.

P

 

 

 

 

a = 40000 thVb,fiyear'

Total U30g Requirements (Metric Tons x 10°)

Year LWR* FBR*%* Total
2000 1.78 (0.0) 1.78
2010 2.90 (.308) 2.90
2020 3.84 (.441) 3.84
2030 4.61 (.397) 4.61
2040 5.20 (.177) 5.20
2050 5.61 (-.218) 5.61
2060 5.85 (-.790) 5.85
2070 5.91 (-1.54) 5.91
2080 5.80 (-2.46) 5.80

 

*
The LWR is considered to recycle all Pu not used by the FBR.

kk
Since FBRs use Pu generated from LWRs, no mined U30g is needed for

FBR fissile requirements. The numbers in parentheses are the equiva-
lent U30g requirements of the Pu in FBRs, if U235 had the same nuclear
properties as Pu in FBRs; the numbers give perspective relative to
fissile fuel use.
l.

G-14

REFERENCES

R. W. Hardie, W. E. Black, and W. W. Little, ALPS, A Linear
Programming System for Forecasting Optimum Power Growth Patterns,

HEDL-TME 72-31 (April 1972).

Alexander Sesonke, Nuclear Power Plant Design Analysis, TID-26241
(1973).

 
APPENDIX H

REPROCESSING COST ESTIMATES

Summary: The bases on which shipping, reprocessing, conversion, and waste
storage costs were estimated are described in this Appendix for LWR, CANDU,
HTGR, and FBR fuel. Flow sheets for reprocessing fuel from each reactor
type were drawn. Five stages in reprocessing were addressed in the flow
sheets: (1) head end, (2) solvent extraction, (3) conversion, (4) off-gas
treatment, and (5) waste disposal. The cost of shipping irradiated fuel
from the reactor to the reprocessing plant was also considered. The
complexity of the five stages of the reprocessing flow sheets were compared
with the AGNS plant, where costs are assumed to be known. Estimates of
cost for CANDU, HTGR, and FBR reprocessing were made on the basis of the
comparison of flow sheets with the AGNS flow sheet for LWR Pu-U fuel. _
Two plant-size bases were used in this study, 5 tonnes HM/day and a plant%j
of sufficient size to service a 50 GW(e) industry of a given réactSfAtype.
For LWR fuel reprocessing a 5 MT/day plant will service a 50 MW(é) industry,
but for the other reactor systems considerably larger or smaller plants

are required. This will be discussed later.

A summary of results of the reprocessing cost estimates on a $/kg HM
basis is given in Table H.l1l. The fuel cycles which have been considered
in Table H.l are summarized in Table H.2. The relationship between
reprocessing capacity and installed electrical capacity is given in

Table H.3. It is noted that 5 MT/day capacity for LWR fuel reprocessing
will support a 50 GW(e) LWR economy. However, because of the high burnup
achieved in HTGRs, a 5 MI/day reprocessing plant will support a 157 GW(e)
HTGR economy. Likewise, because of the low exposure achieved on fuel
operating in CANDU reactors operating on the U-Pu fuel cycle, a 26 MT/day

reprocessing plant is required to support a 50 GW(e) economy.
Table H,1. Thorium fuel cycle study: estimated costs of shipping,
reprocessing, and waste disposal

($/kg heavy metal)

 

Plant capacity

 

 

 

5 tonne heavy metal/day 50 GW(e)
Waste
Reactor (fuel) Shipping Reprocessing Disposal Reprocessing

LWR (U-Pu), AGNS plant 5 221 71 221
LWR (U-Th) 5 222 71 222
LWR (Pu-Th) 5 233 83 233
CANDU (U,Pu) (CR = 1.0) 1.5 210 61 86
CANDU (U,Th) (CR = 1.0) 4 207 61 96
CANDU (Pu,Th) 4 223 74 115
FBR (Pu-U), LMFBR 30 294 85 383
FBR (U-Th) 30 302 111 394
HTGR (U~Th) (CR = .66) 30 622 85 1148
HTGR (U-Pu) 30 631 85 1151
HTGR (Pu-Th) 30 626 85 1153
HIGR (U-Th) (CR = 0.82) 30 622 85 923
HTGR (U-Th) (CR = 0.82)

(partial burning of graphite) 30 483 85 717

 

r

 
Table H.Z2.

H-3

Fuel cycles considered for comparative costs

of shipping, reprocessing, and waste storage

 

 

Reactor Initial Fuel Recycle Fuel
LWR (235y-238y) oxide (Pu-238y) oxide
(2335U-Th) oxide (233y-Th) oxide
(Pu-Th) oxide (233U-Th) oxide
CANDU (235y-238y) oxide (Pu-238y) oxide
(235U-Th) oxide (233yU-Th) oxide
(Pu-Th) oxide (233y~Th) oxide
FBR (Pu-2380) oxide (Pu-2387) oxide
(233y-Th) alloy (233y-Th) alloy
HTGR 235y¢,-ThO, 233y¢,-Tho,
235yc,-238yo, (PU-238y) oxide

(Pu-Th) oxide

233yc,-Tho,

 
H-4

Table H.3. Reprocessing Capacity and
Electrical Capacity Equivalence

 

 

Installed Capacity Reprocessing Capacity
Fuel GW(e) Supported by (tonne/day) to support
type a 5 tonne/day 50 GW(e) of installed
Reprocessing Plant capacity
LWR 50 5
CANDU (U-Pu) 9.7 26
CANDU (U-Th) 13 19
CANDU (Pu-Th) 15.4 16
FBR 80.1 3.1

HTGR 157 1.6

 

 
Scope of the Study

 

This study includes costs for shipping irradiated fuel from the reactor to
the reprocessing plant, reprocessing (including product conversion and waste
treatment), waste shipping to a repository, and waste storage at the

repository.

For the HTGR, it was assumed that the refabrication plant would be on the
same site as the reprocessing plant. Refabrication of LWR, CANDU, and

FBR fuel was assumed to take place at a central facility that served
several reprocessing plants. For these reactors, the product conversion
costs include the cost of making oxides (UQ3 and Pu0,) at the reprocessing
plant and the cost of reconverting these oxides at the fabrication plant
into fuel material that meets the feed requirements of the réfabrication
plant. It was assumed that thorium could be shipped as Th(NO3), solution
and no reconversion penalty was applied. Although the costs of ""tailor-
made" oxides are incurred at the refabrication plant, these costs are

included in reprocessing.

Approach Taken

The Barnwell (AGNS) plant and flow sheet was taken as the base case for
estimation of capital costs. Capital costs of other flow sheets were
estimated relative to the base case. Operating costs were scaled from
capital costs, using estimated factors. A second iteration allowance
was made for plant size, using estimated scaling factors. Most of the
estimates made are of necessity qualitative and subjective, but a deli-

berate effort was made to avoid bias, either intentional or subliminal.
Cost Basis

The initial cost basis for this evaluation was the cost of the Allied-General

Nuclear Services (AGNS) plant at Barnwell, South Carolina. Although actual

‘cost data for this plant are not yet available, the general consensus is

that the complete plant cost will be about $800 million. This total cost

was arbitrarily apportioned among the five major areas of the plant as
H-6

$100 million for product conversiom, $50 million for off-gas treatment, and -

follows: $150 million for headend, $100 million for solvent extraction,

$400 million for waste treatment.

Process Flow Sheets .

 

A process flow sheet was drawn for the AGNS plant to identify the principal
operations of the five major systems of the plant for processing 235U02-
238U02 initial fuel and Pu02a238U02 recycle fuel. Corresponding flow sheets
were drawn for each of the other types of fuel discharged from LWRs, FBRs,
and HTGRs (Figs. H1, H2, and H3). Reprocessing CANDU fuel is similar to
reprocessing LWR fuel so a separate flow sheet was not required. These flow
sheets formed the basis from which the cost comparisons were made. A
deliberate effort was made to be consistent in preparing the flow sheets,

both for systems within a flow sheet and between flow sheets.

 

Cost Estimating Procedures . -

Capital Costs

Each of the five major systems of the process flow sheet for the 'unknown"
plant was compared with the corresponding system of the AGNS plant, and an
assessment was made of the relative complexity of corresponding systems.
This assessment led to a "complexity factor'" that could be used to relate
the cost of an "unknown'" system to the corresponding AGNS system. The
determination of complexity factors was somewhat qualitative, being based
upon considerations of process chemistry, nature and numbexr of operations,
and type of process equipment. Complexity factors are summarized on Tables

H.4 through H.7.

A second consideration in determining the capital cost of the "unknown"
system was the relative capacity of the '"unknown" and corresponding AGNS
system. In this case, we used data from other cost studies,! which have shown

that plant capital costs may be related to throughput by a relationship of the

form . “

 
 

 
 
 
   
    

   

   
   
 

 
 
 
  
     
   

   
 
 

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AMATAL st

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sheyon | tisyp:

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s RusutTion
{Cnav)

 

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| (uiw) l

 

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{ mg:

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[ Barres Rnvavan ]
1 }

 

 

 

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i —— — ——— — —E—— —— — — ——

 

 

 

 

 

 

Phosser Lawveimen Cnmave PooewiTavoe I
U0 (Wiydy= by =)0u =iy P (W0p 2y~ PudCadada
L FLystimaman _I Cantinatiom
UPy = Py Pu (CaOu)* Pely

 

 

 

UFs Buirmany
X e

 

Pala Swivssnny
To Marannicavien

 

 

 

e
|
|
|

 

 

 

TR Ruvewai
{As w0

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Wy (Wa + 1y — Mgy H:’. ot |

 

= [Yawew Frasmen |

 

L Tarwme Smnml

 

   

 

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{ Senovian in Fruany

{ ¥xarvon Concommumml—od Ymvoron Sraman ]

 

 

orr-Gas
Rureass Ta Avmetveaen

 

 

 

 

1 Sonam =

 

[ tasve Rastwnavrew

ecwvany]

 

Ciunu S

Rucvern

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Trvsram Stennes
{riwy

 

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w0

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Ans Tronass

L=

- (Poe)

 

EEa Rawevom
(Cwav

  

Ciat Savvvien
SantimaTiem

   
  

  

 

Tu, T, 0

 
 

Tuasss Disservam
VL= U0 (WOg )
Taoy = Th(nd))e

I, Kr, WOy

      

BDowvaur B

 

 

M1,k

 

L, ke, MOy l .

— — o e

 

Ciar Hurae varn
(Cumsmimarien)

| ICuh Voo tran
CantimaTiom

 

 Varemimaviewm }

WOy WE ALCHSyYy

Fute Drssocation

K 0+ 43 00 (His

 

l Futr Avsvavmany

   
   

[ Sevaur Tavnacvrea

 

 

At Tuonun

 

Preb Anjugrimacy Faur Anustemgny
o inm s Thindpda

 

[ 2ovior Eemavar] [ souios Remavar |

 

 

    
   

 

Senvint Lurnasviom
T U Crean

 
 

Tmareary EurTRALTIHM
ol Th Crveus

     
    
 

 

ThiNODg Srennan
~1% Nupas

 

 

Uls

Ssr mmueay ]
Yo farmsniiavion

THEHOPI4 Prnipication |

(iow Tacuanen) -

C O NV ERSION

 

Tarion Rosswas,

1T Srase lasing Rawum. A
(As Tug0) Mg lwaudy s1awie 1y I_““: Sraanse (i)

[ WOs Jcavasan

    
  
  

 

[X21vium Pruamen ]

 

 

 

[(Sarviom Sroman |

Iopwms Sonprion
Siwvem ZaoLiTE)
Rapow Rpmovar
MoLEcuLAR SigvE)
KaAyFTon Rumoval
Sarpvion i Pepee

  
  
  
 

 

DuconTasd ¥ Qv Gag
Ruvuats To Avonstuint

QFF-GAS TREATME N T

 

 

 

 

 

Sorvaee Raswemmn ]  [Wasts Evaseantiam

 

 
 

   

 

Crnan Bevuany I

I Twvetum Svannes
Redvenn (Wow)

| Wasrs Rarenivany
(LW}

—{ WROy Racovany

 

 

Jo

’ WROy Ruewenn

  
  

 

 

Rvrensnx |

 

 

 

   
   

  

Waste Stenant

ILW, LLW)

WASTE TREATMENT

Fig. H.1l. LWR fuel reprocessing.

 

 

Wetn s Mis Tuathn

 

CwWaere Svonnen |

 
 
 
 

 

Tasy Avsulteunr
Pulwos)s

Fare Assavmaany
LS oy (Hovta

Tasr AttTueuy
ALY T Y

 

[ Serws Bumavar ]

[ Sovim Ramavar ]

i

[ s Ruwavar |

 

Bewveant Laveatran
L% Py Cuun

1%\ Cuom

Bovyeert Latuagtum
Acs Tuessn

=

-y !flmu-]

 

 

[Peew Rasusvmen |

[ Faun Awsptmany |

 

 

[ Senms Rewevan |

[ Serniny Ramavar ]

 

 

 

 

Do Tl T T Sewvtnt Lnvhns v
I Pe T AW Y Crenn
b

 

 

 

Orarara PusciriTanion
Vs (W8 )a = Pu(Caludn

A CimAT g
PulCadeda-=Puoe

 

     
    

ConvEngion
Yo (MO Ube

UOa Brismany
R CANWME

-~ 1% Yusan

  
    

  

‘ou Enenanats

Thinee Frerans

Thindyle Summuuy

 

e RavanmitaTon

 

   

T Rastovas,
(e "My 0Y

  
 
     

A Sraga L
Wy (WY v ly= Ny 1a

 

 

 

st rwm Fuatom )

(Tuiviow Sveasan]

 

[ H6a Je _1

Wy ly Faanvieaw
Awy $STenass Lw)

 
 

 

 

    
   

e

 

   
    
 

Rapow Removai
{MoLrcuar Sieve}

 
  
  

 

Krvyrrow Removaie
Somrvion iu Freow

[Knnoten Concanvusnwu]

   

DaconvTaminares Or-tas
Aeaast To Atemamass

OmML DWG TC-STRE

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Sawwvany l\.:mu]

| Wasta Ewmranarion

 
 
  

 
  

 

Citar Sonwamt
Wxeuevn

| WRoy Rgancin lfi

   
   

Iwtsmiwm St IIA.I]
(niw)

  
  

 
 
  
 

 

wasre RuregrTone
Luuw)

HHNOy Rucavumy

 

 

 

Wasta STenase
(Tiw, LLw)

       

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BY-Th b
N Bomeas

    
 
 
   

 

 

W, RuLnage Fus. Rucuivine]
To Atowsrwans C_Aws CLasmine J

LF_\.}_‘_\_.._ Stomaoun

Sies Mesucviaw
(Crmowd |

   

 

 

 

Svmuh'l_s“g'v e Lues

 

'—mflun\\\t HSopmu Rewmavarl

 

 

   

 

 

 

I iwmaciatan Funs |
ALY IV '.“‘“.“l

 

PuOe Pule
0 18800

l

FuasL Fmibsmunsy

Wi Ruivase Fuse RECuIvimG
To Atmesrmans _Aws Cruamime ]

FutL Stonage l

 

 

 

 

 

   
   

      

(Poov)

Sizw Raputtios Syamusss Grane Ewos
__(Cwery Revositony .

2R {Volonmarion |

 

 

» —-
—— LMy VoLoX IDATIOM HNOy.
1, %y, WO JLRALHING Canciime .
B e 1ty 0L{N0), HuiLs Compacvion 1, Kr WO Pu OL* Py (NI iy Camracyiom _]
Th= ThiNOwe | xesyor=srwuon (e T

 

 

 

 

 

 

 

   

Waste Ruvosivomy

o

 

Fans Asjuarmgur

 
 
 

 
 
   

 
 

DowvanuY Exvnatniown
Aciw Ynanan

I

 

 

 

 

[ lnuu—‘] Feip Avjustmant
2980y (NOy)n ThiWOyle

 

 

[ Sorne Rewwin] [ Sovies Reamevar |
L 1§

Douvenr Butamten

 

 

 

 

S owwewt Tatwacy \.-J

 

 

 

 

 

 

 

 

 

   
      
 

   
   

 

 

 

 

   

 

 

 

 

___ __________________I[______'__.‘__

 

 

 

 

=

Taivwom  Ramovac 1Y Svase [opime Rumovac He, Iy Fixaview
(As "Mr0) ey (N0 +I—=MHy 1 Aus Svtenans

 

   
 
 

 

[ Tritwom Fuxaviow |

| Tritium Stonace |

NO x Scrusewn
Hey(Mdy e

LW Syacn Tovima Ruwewan
Hy (NOpdp, + 1o He 12
Ovr: Gas FiLTmATIAN
looine Somprion
(SiLvem ZeoLive)
Raoon Removar
WMoLEcuLAR Siave
SoRPTION In FrREON

[Ku\v Tow Coucufitnvnu}_‘-' Kavrtow Sieunes I

 
     
   

 

 

Saus Wasre
Svouace (LLW)

 

 

  

 

 

Dacomtaminates I5e-Gas
RULenss To Avemsrunas

]—.—.—_—...._._...

 

T\ Cuere 1% Th Cucnu
0 €
|1 |
Prowwet Cowvangion Th(W0s)e FTonacs
IO (MO )= MITYD, ~ 16 Ngans
YO Bwirmswuy Th(WO3)e Pamiricatian
(lom Exemawcs )
Th(NOy)a Suitmuny
Te RurpmsnicaTIOM
-[— =50 N

 

 

 

 

 

 
  
 
   

        

 

 

 

Soiies ®awevis }— o

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 
 
    
     

 

 

 

 

 

  
 

  

 

   

 

 

 

 

 

 

~———{Servemt Rucovumy] |

 

Wasve Evasountiow Waste HauTRALIZATION
l MpOW = Na¥Nl,y

Waats CaLcimaTion

 

 

   

 

 

 
   

  

    

CLtan SouviLme ' “\\rr‘\un SYsnase
- haw)

Racuinn

 
 
   
 
 

Wasvs RarensTony
(MLVW)

 
  
 

 

 

l YOy Raauns

_ HNOy Rucavenx

Wasts Svonass

 

 

 

(i, Lw) 1["

 

-

Fig. H.2,

 

  

 

 

 

. Seaveant Enn:n.:“ou- |
: AT Puses Cneve
' 2%840g (N Pu (NOS8)
i
E LSo\.\“ Wanovar | EENE T
I t 1 1
! Sowveny Extunivun Sowvant Extaacvion
l | 2wy cuens 2% Py Cnevw
tE X A CTVON
Ly |
Prasvcr Comvaugran Oxavavs Precisitation
1M Y0L (MO )y 1% U0y Pu (M0g)a* PulCa0ada
CarcmaTion
Pul(C10a02™ Puly
PuOr Svirsmant
r - To Revasmic ation
J S\
___[m___fi________ — — —
Tuitiuve Ramovac VAL Svace loov Rumivac He I3 Finavion
(Aa "Wed) Mg ({NOL v 1a* ¥y 1 Ave Srounse
[ Toirium Fruavion] NOx Scrussen | ety
H.(u_o_!h
‘ Taitiom Stonase | % STane lamwe Wemouns
Wy (W0y)z La-=wy 1o
[ore-Gas Fiutnation Sounm WasTe
Stonase (AAN)
Tooime Sanrview
{ Swwven Zasuva)
Kavovan Rumaovar
(Sanrvion in Fugosd
[Xxxrron Comcamznatiom——— Kuvrrom Sromass |
Ducoutraminatuy Ore-Gay
Raisasa To Atenmsrnann

 

 

 

Recoveny

Waste Evarsaavon

{ Wasts Catciuation

I tuniw STouast
(R w)

     
    
    

WAttt MesTravizanion
WalM = NaNl,

 

 

  
 
 

 
   
  

Cluan SoLvewm~
Recytruw ]

 
 

 
 

Waste Ruvositany
(W)

  
    

HWHO3 Rucovany

 

 

 

Watte Svonass | .
(W, wuw) |

FBR fuel reprocessing.

aouNe OWGE 76-3137

 

 
 

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R Tamg) O Pul { Trane)

Tuvtan Mawies
Py Ou Crady [ 178000 (Taena)
PRUT(Brsed | =egoutoan IRLUILES RIS

O T S | s St ) A d

 

1 Tumapiaren Fuso
| Vet b P A I Racvc.n
.

 

 

 

   
 

" ] T " Ridcmivine Ans Svennss Auctwind Ant Brstase
AP lee ooy _
_-[ Twenss & Wavmae ‘Siin Ruewevian Tveaass B Tavs Siie Ry .. T v -
MUty Paavitiesy t___ . (Cavemimay J-—__.! -W‘-{ff‘.\-‘-:"l:jmutt--. [ ‘(Cnu\-.\v::j ™ | - T T : b s,_t:‘:::‘:)“ J'—'— T T,

Puveity Buanius M., %e, Rw 0L, CO
tece

         

T Pmwmany Bunmine MK Yy S Parman: l_‘._‘ o~ M, 1, Ke. Rw, €0,C0,
wwa o C=Cos I T :“‘! e I o C_::'it.e;. - * . ™
Rnvima Fanvicit Sarauavien Luacwiue Panvicen Jusasntian
LU= T M (M, _—— (E_\:\!:'I\.p.v_..n) . Tnoa=Tn{wteis (Bruvniavianw)

 

 

 

 

 

AT .

Laktwiag Panvie
tevehoe P— earniamemy

     
 

 

 

 

 

  

  
  

 

 

 

 

 

 

      

 

 

 

 

 

   

 

 

 

 

 

 

 

 

Bl Wuny Cavomine SiC Wi Cantmine SiL Wwie Cuyvame - 5iC Wure Crupmens
R 1Y tmucy $ula ™ PuOs
.1 %, A=, 56,00
Tusmnn T o =
iy neansamn Iumu?l I Secensant Bynume LB, €0, 80 TS Stoutaty Butarnt
Muts= LU Og C=COr == COn
) W1 R Woa -
s 1 | ] ! - [ e 1 1

| Funs Dissmvnen Tapans Orasorwvian I Fua YT v FusL Disseiyran Yur. Dveveutien Fuse Disterutiew Tuen Dusssnuan Dissowevien
. b, L St Lo} Thou=Th (Hiehy I\ 0 g ARG o (MO Pu D1 Puinbdele A8 yo - $ 98 1) 0g (WOY PuDg == By (WO L9y, O =5 05010y {Wighy Tnds=Th(W0)a
; T ——————— e A 2 —— AT
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Fuge Avsmsramt Frav Avivivanny
SShube (RIn MO L WD TR O

     
    
  

       
    

Feur Asletimeay
P lBuie 3™ U0 (NOy

Fans Anpnevvant
LRuon (Mgl Th (R84)4

       

 

Hoviny Rampyar

Sovant Timacwen
Avs Tnanas

Frun Asaentmamt
Loy (MO0

Boive Pamevar

Faxs AbjusTmeny
S0 (WA
Borwt Resey dn |

i
St Errhatvm TuTLAL T
LHB 139y Cnirm LU Y Gt

 

 

 

 

 
   

Sevent Ravansram
1AE Cine Pumun
Four Avivsnmnr Faur Remrrwans
Th LU0 BESYDL (ML

L Senior Ruvamwarn | [Gevws Revavme | Som Ramavir |
1 l i

Y e That Yo Seurtert Sxtiartyu Soussay Levascoes
T 1y Cutum LU A%y Canauy TS T Cucn

Jouvant Revalvwe
A Crom Funuu

 

 

 

    
   
 
     
 
    
 

 
   
  
   
   

      
  
  

 

Fane Asservecut
Pu NOyde

Pons Ammptugnt
Tu{ Myl

  

 
 

s Anapntmpuy
Wy (WO Y

 

     
 

Ben1us Nanovar

 

 

 

 

 

Sewvawn LrThacrs
TR LYy Cuane

 

 

 

 

 

e
e Tl Bl
i
v
I
fi
z

 

 

 

 

 

 

 

 

 

 

 

V0, (MO Peum aian I ‘(b(‘i\u ransas

(10w Batnsmen) 3 Yumms

  

 
   

UG (NOg )y Tunasrms, Pu (WOyde Tunnsran 100 y0n { M0 Taameman
To Fammuaviewn Ta RAmantatTwn To Wnvasmitnton

 

       
   

 

Pu(WOg)y Taasma S50 (M0n e Pasusrcatnses ThiW0:)ly Sronces
Yo Rasapaicaton 1au Encnrnen ~ 1% Yusne
|‘“uhl&hm |
Ta Rurannisaviow

  
  
   
 

 

IR0 (W Taame s

Tty g Poastruation
To Ravashcatien

(Towt G atuanien)

   
   
  
  

T inty)e Pumrvatum
low Rpanangy

  
   

Wby "
o Rusontcat s

CONVERSION

 

 

 

 

 

 

 

 

 

 

 

 
   
   

    
     
 
     
    
 
   

 

 

 

Ore-Cas Onmavrian
Co=Cop
By S O

W02 Ducomwrenvion
Whe == My Wy O

Orr-Gas Onunatisn On Decomrosimon
9y 0 Hos =Ry W &

Tosum Rasewar
(Laas Loswve)

weine van
{Sven Trowrra)

 

 

  

lesn Rmmsvan
(Luss Zomatad

 

      
    
 

 

 

 

 

 

 

 

 

Tavvapa Ruswse.
.33

 

Wavran Rpasan
{ XMt Puayun)

     
   

 

 

Orr-Gan Fuveenan |

 

 

 

0er-Cas Fuxnsnom

DagsuTamuntes OovGan

Ot tonvammaren Ove-Gas Dacauvisunatoe Qri-Gas
Rmaase To Mmawema Rocnase Ty Mrvasiems Rmanye o Anvasmenan

QEF - 6AS TREATMEMNT

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

L
|

 

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N

[ Traven ] [fea¥unrar]

 

 

 

 

 

 

 

 

      
  

 

 

    
  
  

 

 

 

 

 

 

   
    
  

    
 
    

 

 

   
    
 

 

 

 

 

 

 

    

 

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l (50s ¥ruaven | [Facyme frzavem | 1
l Cunam Soummar Tmvvigm Waste Torvm WasTe Qg WanTe Tt Wasve
I Weesten Stanaes Steunss : Bvennss x Svenndt
. Pi=dy-tirhde v . —===ae ]

' Waste Sonin Whers I Wasra S Warte “Wuoy T e ) i Tomas s

Evasetatiom Raretiveny (Limiim) Evatenatinm Rusetrvenriivmim Iua"::" Lvassnavum Warearrons { 1L 11w

t - . R -
' < &0_."'_\_ m-h:n_.g_-l | wuey Rucenn ] Wasts Cavtimariom |
1 tra Tatea Wasva
l o LSranasy ()
I ot _fn'—--tn Aot Taly w‘n“-'-‘;;;:al ! B 1 Raretivont . .w:;u;s"fw-;;-'l‘ ! Q.\Ta I;u‘!no& |
‘a‘;‘.‘,xm $:iC Vona s (Mrw) ] L_( ww W) I I 31 Woues (RLw} RALL AL I, tMuwn
I . WASTE TREATWMENT
ol o e mmae e Al

L e e e e e — e

Fig. H.3. HTGR fuel reproéessing.

TT-H

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"I'“

Table H.4.
Power System
Reactor type , LWR 238
Initial fuel 5y0,-2*%y0,
Recycle fuel Pu02-**%U0;
Clad or coating Zircaloy-2

Reprocessing Plant Capacity

Heavy metal, tonne/day
Clad or graphite, tonne/day
Equivalent GW(e)

Irradiated fuel shipping
Reprocessing plant

Head end

Solvent extraction

Product conversion

Off-gas treatment

Waste treatment and storage

Total

Waste disposal

Shipping
Repository

(AGNS plant)?

Thorium fuel cycle study:

LWR
23540, -Tho,
233y0,-Tho,
Zircaloy=-2

1.5
50

estimated shipping, reprocessing,
and waste disposal costs for light water reactors

LWR
£yga-Tho:
V0z-ThO2
Zircaloy=-2

50

Estimated costs of shipping, reprocessing, and waste disposal

 

__r_fimumu__.QMKNUng
(10° 3) ($/kg HM) ($/kg HM)

Factor
(Complexity) (size)

—_~Capiral

Op
(10° $) ($/kg HM) ($/kg HM)

erating

Factor ——LCapital _ Op
(Complexity) (size) (10° $) ($/kg HM) (

erati
$/kg HM

 

30
20
10
80

160

N W

61

(1.1) ()
1y (@
(0.7) (1)
1y @
(1.05) (1)

(a) The AGNS plant was taken as a base case
and costs of processing "unknown" fuels were
related to these costs.

(1.1) (1)
(1.15)(1)
(.95) (1)
1 @)
(1.05)(1)

W~ W
. + .
N WO N

w
o

64.3

¢T-H
Table H.5.

Power System

Reactor type
Initial fuel
Recycle fuel
Clad or coating

Reprocessing Plant Capacity”

Heavy metal, tonne/day
Clad or graphite, tonne/day
Equivalent GW(e)

Factor

(Complexity) (size)

Irradiated fuel shipping

Reprocessing plantb

Headend

Solvent extraction
Product conversion
Of f-gas treatment

Waste treatment

Total

Waste Disposal

Shipping
Repository

(1.05)(9.23/6.5)'6
(1) (3.12/5)*
(.9) (3.12/5)°®

(1) (3.12/5)" ¥

(1.05) (9.23/6.5) " >°

Thorium fuel cycle study: estimated shipping, reprocessing,
and waste disposal costs for fast breeder reactors

FBR FBR

Pu0 —238U0 233U-Th alloy
Z 53 2 233

Pqu- UO2 U~Th alloy

stainless steel stainless steel

2 (5) 2 (5)
3 (7.5 3 (7.5
50 (125) 50 (125)

Estimated cost of shipping, reprocessing, and waste disposal

Capital Operating Factor Capital
__rra e
(10” §) (S$/kg HM) ($/kg HM) (Complexity) (size)  (10° $) ($/kg HM)
30 30

194 (258) 62 (52) 25 (21) (1.1)(9.23/6.5)"° 204 (270) 65 (54)
85 (100) 27 (20) 8 (6) (1) (3.12/5) " 85 (100) 27 (20)
68 (90) 22 (18) 9 (7) (.8)(3.12/5)"° 60 (80) 19 (16)
42 (50) 13 (10) 4 (3) (1.05) (3.12/5)" > 44 (53) 14 (11)
475 (560) 152 (112) 61 (45) (1.1)(9.23/6.5)° 32 497 (587) 159 (117)
864 (1058) 276 (212) 107 (82) 890 (1090) 284 (218)

9.5 11

75 100

Ayalues given in parentheses are for alternate plant capacity of 5 tonnes heavy metal/day.

bCapital and operating cost given in parentheses are for alternate plant capacity of 5 tonnes heavy metal/day.

@

Operating
($/kg HM)

26 (22)

8
8
4
64

110

(6)
(6)
(3
(47)

(84)

71-H

 

 
Table H.6. Thorium fuel cycle study: estimated shipping, reprocessing,
and waste disposal costs for high temperature, gas cooled reactors

Power System

HTGR HTGR
Reactor type
Initial fzzl :::Ucz(Triao)-ThOz(Bino) 225yc, (Trisn)-22"UC, (Bise)
UCz2 (Triso}-ThO; (Biso) Pu0z (Triso) -2 **U0, (Biso)
Recycle fuel raphi
Clad or coating graphite graphite
Reprocessing Plant Capacity?
Heavy metal, tonne/day 1.6 (5) 1.6 (5)
Clad or graphire, tonne/day 18.5 (58.1) 18.5 (58.1)
Equivalent GW(e) 50 (157.2) 50 (157.2)

Estimated costs of shipping, reprocessing and waste disposal

21

HTGR
Pu0; (Triso)-ThO; (Biso)
3UC; (Trise)-ThO; (Bisa)

1.6
18.5
50

(5)

(58.

1)

(157.2)

 

Factoer Capital ’ Operating Factor Capital
(Complexity) (size) (10" §)  (S/kg RM) ($/kg HM) (Complexity) (size} (10" §) (5/xg W1
Irradiated fuel shipment 10 30
Reprocessing plant
Headend (1.8)(20.1/6.5):6 530 (1056) 331 (211) 132 (B4) (1.8)(20.1/6.5)*% 530 (1056) 331 (211)
Solvent extraction (1.25)(1.6/5):33 84 (125) 52 (25) 16 (7.5)  (1.2)(1.6/5):%3 67 (120} 42 (24)
Product conversion (.10)(1.6/5): 33 7 (10) 4 (2) 2 (.8  (.4)(1.6/5)-8 20 (40) 12 (8)
Off-gas treatment (1)(20,1/5)- 3> 81 (121) 51 {(24) 15 (7.2)  (1)(20.1/5)-35 81 (121) 51 (24)
Waste treatment (1.05) (20.1/6.5)-35 623 (931) 389 (186) 156 (74) (1.05)(20.1/6.5)-35 623 (931) 389 (186)
Total 1325 (2243) 827 (44%) 321 (173.5) 1321 (226B) 825 (453)
Waste disposal
Shipping 10 10
Repository 75 75

8values given in parentheses are for alternate plant capacity of 5 tonnes heavy metal/day.

bCapital and operating costs given Ln parentheses are for alternate plant capacity of 5 tonnes heavy metal/day.

Operating
{5/kg HM)

132 (84)
13 (7.2)
4.8 (3.2)
20 (9.6
156 (74)

325.8(178)

Factor Capital

(Complexity) (size) (10° §)

(1.8)(20.1/5.53-6 530
(1.25)(1.6/5) 33 84
(.25)(1.6/5)+6 13
(1)(20.1/5)+35 81

(1.05)(20.1/6.5)-3% 623

1331

(1056)
(125)
(25)
(121)
(931)

(2258)

($/kg HM)

3N

331
52
8
51
389

831

10
75

(211}
(25)
(5)
(24)
(186)

(451)

Operating
(§/kg EY)

132 (84)
16 (7.5)
3 ()
15 (7.2)
156 (74)

322 (176.7)

ST-H
Power System

Reactor type
Initial fuel
Recycle fuel
Clad or coating

Reprocessing Plant Capacity®

Heavy metal, tonne/day
Clad or graphite, tonne/day

Equivalent GW(e)

Irradiated fuel shépping

Reprocessing plant
Headend

Solvent extraction
Product conversion
Of f-gas treatment
Waste treatment

Total

Waste disposal

Shipping
Repository

Table H.7.

Thorium fuel cycle study:

estimated shipping, reprocessing,

and waste disposal costs for heavy water moderated reactors

Factor

(Complexity)(size)

(1) (27.4/6.5)5
(1) (25/5)+35
(1)(¢25/5).6
(1)(25/5)-35
€1)(27.4/6.5)-33

CANDU
235, _238y0,

Pu0,-* %00,

Zircaloy

25.7 (5
2.43 (0.48)
(9.73)

50

Estimated costs of shipping, reprocessing and waste disposal

Capital Operating
(10° §) ($/kg HM) ($/kg 1)
1.5
360 (135) 1l4.4 (27) 5.8 (11)
180 (100) 7.2 (20) 2.2 (6)
260 (100) 10.4 (20) 4.2 (8)
90 (50) 3.6 (10) 1.1 (3)
660 (377) 26.4 (75) 10.6 (30)
1550 (762) 62 (152) 24 (58)
7.4
54

CANDU

23 5U02-Th02
2330, -ThO,

Zircaloy

19.2
1.82 (0.48)
(13

50

Factor
(Complexity) (size)

(1.1)(21.8/6.5)+% 340
(1) (20/5)+35 160
(0.7)(20/5):6 160
(1)(20/5)-35 80
(1.05)(21.8/6.5)-33 640

1380

Bvalues given in parentheses are for altermate plant capacity of 5 tonne heavy metal/day.

(5)

(135)
(100)
(70)
(50)
(396)

(751)

Capital
(10" $})

($/kg HM)

bCapital and operating costs given in parentheses are for alternate plant capacity of 5 tonnes heavy metal/day.

27)
(20)
14)
(10)
(79)

(150}

Operating
($/%g HM)
6.8 (10.8)
2.4 (6)
3.2 (5.6)
1.2 (3)

13 (31.6)
27 (57)

“CANDU
Pu02~ThO3

233110, _Tho;

Zircaloy

16.2 (5)

1.53 (0.48)
50 (15.4)

Factor
(Complexity) (size)

(1.1)(17.5/6.5)-% 300 (149)

€1.15) (16/5)- 35 170 (115)
(.95)(16/5):6 190 (95)
€1)(16/5)-33 a (50)

8
(1.05)(17.5/6.5)-33 590 (396)

1330 (805)

&4

e
oNun = O e
Do e e
O W o

2

B.6
65

Capital
(10° 8)  ($/kg HM)

(30)
(23)
(19)
(10)
9

(161}

Operating

($/kg EM) T
|
=
o

7.5 (12)

3.2 (6.9)

4.8 (7.6)

1.5 (1)

15 (32)

32 (62)

 
H-17

Cost of "unknown" _ | throughput of "unknown"] n
cost of AGNS throughput of AGNS

The expression in brackets above, the capacity factor, was calculated for
each section of the "unknown'" plant, and thé estimated cost was found by
multiplying the appropriate AGNS cost by the product of the complexity
and capacity factors. The scaling factor, n, was taken to be 0.6 for
headend and product conversion and 0.35 for solvent extraction, off-gas
treatment, and waste treatment. Throughput included the total amount of

material passing through a given section of the plant, that is, heavy metal

plus cladding or matrix material.

Capital costs were determined for two plant sizes: (1) a processing plant
that treats 5 tonnes of heavy metal per day and (2) a plant that treats the
fuel discharged from reactors that produce a total of 50 GW(e). Plant opera-

tion of 300 days/year was assumed.
Annual Capital and Operating Costs

An amortization rate of 307 per year was used to determine annual capital
costs. Annual operating costs were calculated for each major area of the
plant and were taken as either 40% or 307 of the corresponding annual
capital cost. Operating costs for the more labor intensive and higher
maintenance areas (headend, product conversion, and waste treatment) were
computed using the 40%Z rate, whereas, costs for the more conventional
chemical plant operations (solvent extraction and bff-gas treatment) were

calculated at 30%.

Shippin

Considerable research and development has gone into the study of shipping
irradiated fuel elements, and reliable cost data are available. Shipping
costs for the several fuels of this study were estimated from the data

reported in WASH-1099,2 with escalation to reflect current costs.
H-18

Very little is known, however, about the costs of shipping and storage of
nuclear wastes. Preliminary shipping costs3 have been reported for HTGR
wastes. Waste shipping costs for other fuels were estimated using the HTGR
costs as a basis, with consideration given to types and quantities. Shipping

costs include an allowance for worn-out equipment.

Waste Storage

Waste storage costs were estimated by the same procedure as waste shipping
costs. The cost data of the HTGR fuel cycle3 were taken as a basis, and

costs for other fuels were related to these wvalues.

 
H-19
REFERENCES FOR APPENDIX H

Staff of the Oak Ridge National Laboratory, Reactor Fuel Cycle Costs
for Nuclear Power Evaluation, WASH-1099, pp. 60-80 (December 1971).

Staff of the Oak Ridge National Laboratory, Reactor Fuel Cycle Costs
for Nuclear Power Evaluation, WASH~1099, pp. 99-117 (December 1971).

HTGR Fuel Recycle Planning Team, Summary Program Plan - Alternate
Program for HTGR Fuel Cycle (Draft), p. 148 (Apr. 11, 1975).
3,

3

 
Appendix I

FABRICATION AND REFABRICATION COST ESTIMATES

Summary: The method used to estimate fuel fabrication and refabrication
costs for LWR, CANDU, HTIGR, and FBR systems is similar to that previously
described for reprocessing (see Appendix H). Flowsheets were developed for
each type of reactor fuel, and the complexity and specialized equipment
requirements compared. Since no commercial scale facility exists for remote
fuel fabrication, and the cost data for fresh fuel fabrication are not

generally available, previous ORNL estimates for LWR fuel fabrication were

‘updated, and used as a comparison base. A summary of the results of this

study is contained in Table I.1 for plants with 2 MT HM/day output. The
factors listed in Table I.2 can be applied directly to the $/kg values appear-

ing in Table I.1 if estimates for other size plants are desired.

Fuel Fabrication Cost Estimates

 

The large variety of fuel materials and fuel element designs considered in
this study together with the limited time precluded a formalized estimation
procedure such as that done previously.la2 However, one of the cases from
these early studies formed the basis for the reference base case for metal
clad cylindrical fuel rod types. This LWR (PWR) case from FABC@ST 9 provided
the appropriate distribution of cost elements under the categories of
Capital, Hardware, and Operation. The costs in each category were escalated
from the 1966 data by assuming a 107 per year inflation rate, as was done in
a previous study,3 and adding both capital (50%) and operating (307)
increments to incorporate the features for current or proposed requirements
for total liquid recycle, scrap reprocessing, and solid waste treatment,

particularly transuranic waste."

With this as a basis, the fabrication process outlines given in Figs. I.1
through 1.4 were used to make a relative factorial estimate for incremental
features in each category of cost. The hardware cost factors were based on
available fuel element design data and evaluation of three increments:

cladding (with end caps), fuel rod internal component complexity, and
 

 

 

Table I.1. Estimated Fabrication Cost Comparison?
Reactor Fuel Relative_Cost Factors Esgizi:ed
Type Material Capital Hardware Operating Total ($/kg)b
PART A
LWR (PWR)  (235u-u)o, 0.33 0.38 0.29 1.00 150°
(Pu-U0)0, 1.49 0.38 1.45 3.32 500
(235U-Th)0, 0.50 0.42 0. 44 1.36 200
(233y-Th)0, 1.98 0.38 1.45 3.81 570
(Pu-Th)0, 1.49 0.38 1.53 3.40 510
CANDU Normal UO, 0.33 0.09 0.11 0.53 80
(Pu-U)0, 1.49 0.09 0.50 2.08 310
(233y-Th)o0, 1.98 0.09 0.50 2.57 390
(Pu-Th) 0, 1.49 0.09 0.53 2.11 320
FBR (L.M.) (Pu-U)0, 3.19 0.58 2.10 5.87 880
(Pu-U)C 2.68 0.37 1.66 4.71 710
233y_Th 2.73 0.35 1.60 4.68 700
FBR (Gas) (Pu-U)O, 3.19 0.90 2.29 6.38 960
(233y-Th)O, 4.55 0.90 2.40 7.85 1,180
(Pu~Th) 0, 3.64 0.90 2.40 6.94 1,040
PART B
HTGR 23540,-ThO,  0.26 0.42 0.32 1.00 4004
233yCco-Tho, 1.21 0.42 0.95 2.58 1,030
23500,-U0, 0.26 0.32 0.32 0.90 360
Pu0,-ThO5 1.21 0.42 0.94 2.57 1,030

 

a * - -
All cost comparisons are relative to the given base case factors.

b1977 dollars assumed for total kilograms of heavy metal product with a
plant output of 2 metric tonnes per day and 260 full operating days per year
(520 MT/year). .

CBase case for metal clad fuel rods based on FABCOST 9 estimates

(A. L. Lotts et al., A/CONF, 49/P/062, 1972) escalated to 1977 with additions
for current scrap and waste treatment requirements.

dBase case for all HTGR (Prismatic Fuel Element) cases based on data in
"Summary Program Plan, Alternate Program for HTGR Fuel Recycle,'" April 11,

1975, Draft.

 
Table I.2.

Fabrication Cost as a

Function of Processing Rate

 

 

Rate Cost
(MT HM/day) Fraction
0.5 1.53
1.0 1.23
2.0 1.00
3.0 0.90
4,0 0.84
5.0 0.79
6.0 0.76
>7.0 0.73

 
PuO2 POWDER

U0 POWDER-11
UO2 POWDER—‘ i

ThO2 POWDER
f,--rnoa POWDER
|

4 rPuOg POWDER

 

 

BATCH BLENDING

 

 

v .

 

 

 

 

 

 

 

SAMPLES @—

 

 

 

 

 

 

 

 

 

MILLING -
BINDER
¥ ;
LOT BLENDING
v
AGGLOMERATION |
v
GRANULATION

 

 

 

 

y

 

i

 

 

PELLET PRESSING

 

 

4

 

BINDER REMOVAL

 

 

 

 

 

 

 

 

 

 

 

Zr2 OR Zr4
TUBING

 

 

INSPECTION

CLADDING TUBE

 

 

Y

 

CLEANING

AND PICKLING

 

 

¥

 

 

y

 

 

AUTOCLAVING

 

 

 

 

 

FINES
OXIDATION/REDUCTION GRINDINGJ

 

Y
SINTERING
Y
RESCCORV‘EPRY GRINDING
Y
sampLEse— | PELLET
Y

 

 

PELLET STACK

FORMAT

INSPECTION

ION AND

 

\

 

 

 

 

END

r-PLUG

 

BOTTOM
END PLUG

 

 

Y

 

INSPECTION

 

 

 

 

 

 

¥

 

 

FUEL ROD
LOADING

 

 

INTERNAL

 

 

 

 

 

 

#

4

y ‘ HARDWARE

UPPER PLENUM
COMPONENT
INSERTION

ORNL-DWG 76-5070

P

 

TOP END
DECONTAMINATION

 

 J

 

FUEL OFF-GASSING

 

y

 

 

TOP END PLUG
INSERTION
AND WELDING

 

 

\

 

 

HELIUM LEAK TEST

 

 

Y

 

FUEL ROD
DECONTAMINATION

 

 

)

 

 

FUEL ROD
INSPECTION

 

 

 J

 

 

FUEL ELEMENT
ASSEMBLY
AND INSPECTION

 

 

 

 

Fig.

I'l.

LWR oxide recycle fuel element fabrication.

 
CRNL-DWG 76-5072

   
  

        

 

 

 

 

     
   

   
     
   

 

 

 

 

    
  
  
      
    
     

     
     

 

   
 

 

 

 

  

 

 

 

 

  

 

 

 

 

    
   

  

     
     
    

 

 
   

 

 

 

 

 

 

 

 

 

 

 

   
   

 

  
  

 

 

 

 

 

 

 

 

 

 

= o e == ThQp = e 233 TYPE 36
I 3 % STAINLESS STEEL
BINDER | UO, POWDER CLAD TUBE
CLADDING TUBE FUEL STACK
BATCH BLENDING INSPECTION OFF-GAS
END PLUG—‘
AGGLOMERATION MILLING
_ BOTTOM END PLUG FUEL PELLET
INSERTION LOADING
SAMPLES st BINDER INTERNAL AND WELDING
GRANULATION HARDWARE
LOT BLENDING l UPPER AX(AL
” b e OWER GAS PLENTY BLANKET LOADING
uw COMPONENT
g PELLET PRESSING INSERTION INTERNAL
5 AGGLOMERATION : r HARDWARE
& L UPPER GAS PLENUM
] COMPONENT.
& BINDER REMOVAL INSERTION
@ GRANULATION
@ TOP END
SINTERING 2 DECONTAMINATION
PELLET PRESSING g
&
o
OFF-GASSING |
BINDER REMOVAL OF FUEL ROD J
&
PELLET 3
TOP EN G d
INSPECTION SINTERING INserr i T
AND WELDING
PELLET STACK
FORMATION AND ‘
INSPECTION 1 A HELIUM LEAK TEST
SCRAP
RECOVERY
PELLET STACK PELLET LOWER AXIAL FUEL ROD
OFF-GAS INSPECTION BLANKET LOADING DECONTAMINATION
h
PELLET STACK FUEL ROD
FORMATION AND INSPECTION
INSPECTION

 

 

 

 

 

1 PIN WIRE WRAP

 

 

FUEL ELEMENT
ASSEMBLY
AND INSPECTION

 

 

 

Fig. I.2. FBR oxide fuel element fabrication.

6-1
SINTERING
AID{~2% Ni)

DER

BIN
(~1*% Carbowax)

 

    
    
 

   

CARBOWAX

CONFIRMATION BINDER

PRESS BRIQUETS
‘—Ar (vocuum)fi

PRESS BRIQUETS

 
 

 

 

 

BATCH CARBOTHERMIC BATCH CARBOTHERMIC
REACTION REACTION
(1750-1950°C) (1750-1950C)

 

 

 

 

 

EXCESS CARBON
REMOVAL

EXCESS CARBON
REMOVAL

 

SINTERING
AID (~2% Ni)

BINDER
{~1% Carbowax)

 
  
   
  
     
   

 

MILLING

<
MILLING | Q
3
=2
=

MILLING AID
DISTILLATION

 

    
  

GRANULATION

PELLET PRESSING

BINDER REMOVAL
(300°C)

GRANULATION

 
   
 
 
  
   
  
 

PELLET PRESSING

     

BINDER REMOVAL

Eil
d

    
 

  
     
   
  

SINTERING
{~20004C)

SINTERING
{1S00-1700°C)

 

  

  
  

   
  
 
   
 

PELLET
INSPECTION

PELLET
INSPECTION

 

 

 

 

 

 

 

 

 

 

 

 

PELLET FRAGMENT
SCREEN TUBE

TYPE 316
STAINLESS STEEL
CLAD TUBE

 

INSPECTION

CLEANING

 

 

CLADDING TUBE
INSPECTION

 

ORNL-DWG 76-5073

 

PELLET FRAGMENT -

 

 

 

BOTTOM END PLUG
INSERTION
AND WELDING

 

 

INSPECTION
AND CLEANING

 

v

= SCREEN FUEL

SCREEN LOWER AXIAL
BLANKET LOADING

|

PELLET FRAGMENT

 

 

 

 

 

 

LOADING

|

PELLET FRAGMENT-

 

 

 

SCREEN UPPER AXIAL
BLANKET LOACING

 

 

 

 

 

SODIUM

 

SODIUM BOND
MATERIAL
PURIFICATION

SODIUM SLUG
FABRICATION

WEIGHING
AND INSPECTION

 

 

 

     
 
     
   

 

   
  

  
 

INSERT
SODIUM BOND SLUG

 

 

 

 

 

 

 

 

GRINDING -
SCRAP
PROCESSING
PELLET STACK PELLET STACK
FORMATION AND FORMATION AND
INSPECTION INSPECTION
Fig. I.3. FBR U-Pu Carbide fuel element fabrication.

HELIUM LEAK TEST

il

 

LOAD FRAGMENT
SCREEN INTO
CLADDING

 

 

 

  
 
 

 
  

THERMAL/VIBRAT IONA
BONDING

SODIUM BOND
INSPECTION
INTERNAL
HARDWARE

FISSION GAS PLENUM
HARDWARE {NSERTION
CLEAN TOP END
Of CLADDING

TOP END PLUG
INSERTION
AND WELDING

 
 
 
 

 

 
 
  
   

 

 

 

 

 

 

FUEL RCD
DECONTAMINATION
FUEL ROC INSPECTION

FUEL ELEMENT
ASSEMBLY
AND INSPECTION

 

 

 

 

 
 

 

CALCIUM

CHLORIDE  Tho, UF, PuFs

CALCIUM POTER POWDER  CALCIUM POWDER  CALCIUM
REDUCTION REDUCTION

l 975 —] L 650 °C | REDUCTION

 

 

 

  
  

V-20 TYPE 316
Ti  STAINLESS STEEL
CL'AD CLAD TUBE

CLADCING TUBE

  

INSPECTION

 

 

 

 

 

 

 

 

1 1 ACETIC BOTTOM
§ aco END PLUG
| CRUSH —I L WASH WASH BOTTOM END LG
INSERTION
wvos—y | ! i SCRAP AND WELDING
L LEACHING I L DRY U POWDER ] | DRY Pu POWDER
WASTE ] INSPECTION
TREATMENT - AND CLEANING

DRY THORIUM
POWDER

  

'

 

 

 

l T
[:RESS BRIQUETS —| L BLEND j

:
L ARC MELT j l PRESS BRIQUETSj
; —

| CHILL CAST |
BLANKET SLU%1 ARC MELT *I

TRIM SLUGS

:

.

  

CHILL cAsT
ALLOY SLUGS

 

LINSPECT SLUGS | L TRIM SLUGS }—.

:

  

INSPECT SLUGS

LOAD LOWER AXIAL

 

 

Fig. I.4.

BLANKET SLUGS

LOAD FUEL SLUGS

LOAD UPPER AXIAL
BLANKET SLUGS

    

  
     
 
    
   

BLEND

REJECTS

 

SODIUM

 

SODIUM BOND
MATERIAL
PURIFICATION

SODIUM SLUG
FABRICATION

WEIGHING
AND INSPECTION

 

 

 

      

    

 

 

 

INSERT
SODIUM BOND SLUG

 

 

 

 

ORNL-DWG 76-5071

  

THERMAL /VIBRATIONAL
BONDING

 
  
 
    

SODIUM BOND
INSPECTION

  
  
  

INTERNAL
HARDWARE

FISSION GAS PLENUM
HARDWARE INSERTION

 

CLEAN TOP END
OF CLADDING

 

TOP END PLUG
INSERTION
AND WELDING

 

 

 

HELIUM LEAK TEST

FUEL ROD
DECONTAMINATION

FUEL ROD INSPECTION

 

 

FUEL ELEMENT
ASSEMBLY
AND INSPECTION

 

 

 

FBR 233U-Th metal fuel element fabrication.

L-1
I-8

assembly components complexity. All capital cost factors included
increments for buildings and equipment. A high level of mechanization
was assumed for equipment, but the degree of automation varies as do
the building costs in accordance with (1) the mode of operation from
contact through moderate shielding, process step containment to signifi- .
cant shielding, and total process step containment, and (2) the account-

ability and safeguard considerations depending on fissile material and

enrichment. Operating costs were derived from six weighted incremental

costs covering cladding preparation, fuel preparation, rod loading element

 

assembly-inspection-packaging, scrap recovery, and waste treatment. Although
reference was made to some previous studies and cost estimations in developing

the factors estimated for the various increments in each category, no attempt

was made to normalize any case to such studies for metal clad fuels. A

separate base case was derived for the unique configuration and fuel form

of the HTGR reactors utilizing a recent ERDA task force study draft.”

The resulting relative fabrication costs comparison is presented in Table I.1.

The precision of any category is probably less than for the total factors, . '
particularly when one considers the options of trading between capital and

operating costs that are available to any commercial venture. The absolute

cost estimates are all given in 1977 dollars and are all for a common produc-—

tion rate plant of 2 metric tonnes per day of heavy metal product with a

capital fixed charge rate of 307 assumed. Within the accuracy of these

estimates (*¥25%), the scaling factors for plant capacity are probably

equivalent to those presented in the Geneva 1972 paper of Lotts et al. from

the FABCPST 9 calculations. Thus a scaling factor can be derived from

Table I.2.

The cost estimates are based on a given fuel element design for each
reactor. No attempt has been made to judge the distribution of various
fuel elements since a distribution of types within a given reactor is
feasible in some instances and is therefore a design variable available

to the core design and fuel management scheme.
REFERENCES FOR APPENDIX I

A. L. Lotts and T. N. Washburn, ''Use of Computer Codes in Estimating
Fuel Element Fabrication Costs," Nucl. Appl., 4, 5: 307-19 (May 1968).

A. L. Lotts, T. N. Washburn, and F. J. Homan, FABCZST 9, A Computer
Code for Estimating Fabrication Costs for Rod-Bundle Fuel Elements,
ORNL-4287 (August 1968).

A, L, Lotts, T. N. Washburn, L. Giller, H. H. Klepfer, and W. H. Layman,
"Status of Thermal Reactor Fuel Manufacture in the United States of
America," Peaceful Uses of Atomic Energy, Vol 8, 1972 (14 Conf.
49/p/062).

Proposed Amendment, 10CFR20 (FR Vol 39, No. 178, September 12, 1974).

Summary Program Plan, Alternate Program for HTGR Fuel Recycle, Draft,
April 11, 1975.
APPENDTIX J

INSTITUTIONAL CONSIDERATIONS

This study has shown that adoption of thorium cycles in thermal
reactors results in better ore utilization than does use of the
uranium cycle. At the same time, if Fast Breeder Reactors (FBRs)
are commercialized on planned schedules, their use with the
uranium cycle gives substantially better ore utilization in a
growing nuclear economy. Thus, development of thorium fuel
cycles corresponds to developing a contingency position for the
case of a delay in FBR introduction. Further, thorium fuel cycles
provide flexibility in the future if FBRs are introduced on
schedule. If anticipated trends for relatively low nuclear
electricity growth hold, and the breeder can be commercialized
on the present ERDA schedule, the contingency position is not
necessary. However, if nuclear electricity demand accelerates
and/or the breeder is delayed significantly, then a contingency
position is prudent. Advocates of the LWR-LMFBR scenario might
argue that any money spent on contingency fuel cycles could be
better utilized on the FBR program to increase the probability
of meeting the present schedule. Those who advocate development
of a contingency position think it unwise to risk everything on
one system which may not be delivered on time. Both arguments
have merit; so deciding between them requires a realistic assess-

ment of the costs, risks, and benefits.
There is a school of thought which believes high gain converter
reactors can replace FBRs in the nuclear picture, and provide

the means to generate electricity until more advanced systems
(fusion, solar) are commercially available on a large scale.
Whether this is practical depends very much upon the nuclear
power growth, the amount of natural U;0g available at reasonable
costs, and the introduction schedule of the advanced systems.
Based on present estimates, FBRs are needed to maintain antici-
pated nuclear power growth. However, introduction of high gain
converters (with conversion ratios approaching unity) does permit
a substantial increase (relative to LWR use alone) in the nuclear
power level which would be practical for the case of a substantial
delay in the commercial use of FBRs. The results obtained here
indicate that high priority should be given to the FBR, but that
a contingency position can and should be developed which requires

development and application of the thorium fuel cycle.

Use of thorium fuel cycles in thermal reactors will require the
development of economic fuel recycle technology. Utilities will
be reluctant to invest in the higher fuel inventory of thorium
cycles unless there is a demonstrated, economic fuel recyéle
technology available to them. The above is particularly true

of thorium-cycle LWRs and HWRs (HTGRs can store fuel for a number
of years more economically then can the other concepts, but would
require fuel recycle about 10 years after introduction). Further,
early introduction of the thorium fuel cycle would require use

of present reactor designs. Thorium fuel cycle development would
be expedited by close collaboration with reactor vendors as well

as with utilities.

 
Introduction of HWRs and/or HTGRs into the U.S. economy would
require substantial investment in those systems. HWRs would
have to meet U.S. safety, safeguards, and environmental regu-
lations, and what influence they would have on the present
CANDU-type design is not known at this time. Further, the
estimated capital investment required in heavy water facilities
would be very large, and greater than the cost of uranium
enrichment facilities which HWR introduction could displace.
HTGRs would require substantial investment in component
development and testing, basic R & D, and "first of a kind"
type costs. Thorium fuel cycle R & D would be required for all
thermal reactors employing that cycle, but would be greatest for
the HTGR. However, for operation on the thorium cycle, LWRs
and HWRs would economically require commercial fuel recycle

facilities earlier than would HTGRs.

Commercialization of U/Th or Pu/Th fuel cycles will introduce
safeguards requirements on fuel fabrication and refabrication
facilities which are not currently in force for manufacture of
low-enriched uranium fuel. The full costs associated with such
safeguards are not yet known, but are anticipated to be high.

The extent of thorium cycle utilization may be curtailed by

the need to produce plutonium for FBRs, and therefore the expense
associated with installation of safeguards may not be justified

in the eyes of the fuel vendors.

The reference nuclear development scenario for the U.S. calls

for Light Water Reactors (LWRs) to provide power and produce
plutonium to be used in LMFBRs. According to the simple model
presented in Appendix P, about 607 of the plutonium produced in
LWRs over the next 30 years must be stockpiled for LMFBR
inventories. If thorium fuel cycles were introduced in LWRs, the
extent of introduction would be constrained by the requirement

to stockpile plutonium. The investment in R & D needed to
commercialize thorium cycles in LWRs may not be justified in
view of the modest improvements over the uranium cycle with
uranium and plutonium recycle and the constraints imposed by

the need for plutonium for use in Fast Breeder Reactors.

 
APPENDTIX K

STUDIES AND PROGRAMS REQUIRED TO "AMERICANIZE"
THE CANDU SYSTEM

Any planned program supporting CANDU reactor development in the
U.S. should be on the basis that it leads to introduction of HWRs
which can compete economically with other reactor systems. Primary
economic features which favor CANDUs are their low fuel cycle costs
and low separative work requirements for uranium enrichment. However,
the cost of recovery of Pu from CANDU spent fuel appears relatively
high per unit gram of fissile, such that it does not appear economically
desirable to recover Pu from natural-uranium CANDUs. This probably
would not be the case if slightly-enriched uranium-fueled HWRs were
employed. Thus, introduction of HWRs into the U.S. might better be
based on use of slightly-enriched uranium-fueled systems. This
implies fhat in addition to studies involving estimates of HWR
capital costs in the U.S., as well as the determination of U.S.
licensing requirements and associated economic implications, an
associated program of fuel development might be needed to insure that
slightly enriched uranium fuel will perform as required. 1In addition,
a fuel recycle R&D program for HWRs would be required, involving both
the uranium fuel cycle and thorium cycle. Also, HWRs do add a
requirement for large quantities of heavy water. While the technology
of heavy water production is simpler than that of uranium isotope
separation, and the required long term separation capacity is limited,
the initial capital investment for heavy water production in an HWR
economy appears higher than that needed for uranium enrichment plants in
an LWR or HTGR economy.

An advantage of CANDU reactors is that they are now being built
and are operating successfully, 'Thus, they presumably could be

introduced in the U.S. fairly readily once the licensing and capital
K-2

costs requirements of HWR systems in the U.S. are resolved satisfactorily.
Associated work would involve ERDA, NRC, national laboratories, A-E's,
and Canadian support. This effort would require a detailed reactor
design and associated safety analysis studies. The cost of such work
would be dependent on the information available from the Canadians and
the studies required as the work progressed; a minimum effort would
require millions of dollars.

Based on the present type designs of CANDU reactors, needed
research and development work would emphasize detailed evaluation of
core performance under various fueling conditions, extensive fuel
recycle development activities, and fuel irradiation testing. Primary
areas are fuel reprocessing, fuel refabrication, and fabrication of
fresh fuel, with emphasis on technology development and demonstrationm.
Irradiation testing would involve slightly enriched uranium fuel as
well as 23°U/Th and Pu/Th fuels. It is anticipated that an HWR fuel
recycle development program involving Th/Pu and Th/U fuel cycles
would cost overall about $150 million (this assumes that the recycle
of U/Pu fuels from LWRs has been successfully developed and is used
as base technology). The above does not include costs for a demonstra-
tion facility, which could add about $500 million to program costs.
Additional work would include fuel development and testing and
associated postirradiation evaluations costing about $30 million,
and detailed reactor design and reactor physics analyses associated
with fueling evaluations costing about $20 million. Table K1 provides
a more detailed tabulation of estimated research programs and studies
needed to support and justify HWRs in the U.S.; included are estimates

of time required to complete such work and estimates of the cost.

 
K-3

Table K1. Research Programs and Studies Needed to

Support and Evaluate HWRs in the U.S.

 

Activity

Estimated Estimated

Time
Required
(yr)

Cost

($k)

('76 $'s)

 

Research Programs

1.

Reprocessing systems and chemistry for:
a. Zr clad Th0,-235U0, fuel
b. Zr clad ThO,-Pu0,-233U0, fuel.

Decladding techniques for Zr clad fuel
(needed because Zr complicates reprocess-
ing of Th),

Reinvestigation and optimization of the
Thorex and Zirflex processes.

Determination in detail of the process
differences between fabrication of 233U/Th
fuels relative to Pu/U fuels.

Investigation of means to reduce the
positive reactivity void coefficients in
CANDU designs.

Determination of the extent of operational
flexibility of CANDUs to meet U.S. utility
requirements.

Updating of Th, 233y cross section
measurements and evaluations.

Studies

1.

Comparison of capital costs of CANDU and
LWR systems on same basis, including plant
modifications for CANDUs to bring them
into compliance with U.S. standards and
regulations.

Determination of importance of nuclear
growth rate in the competitiveness of
CANDU-Th vs CANDU-Pu recycle modes.

won

50,000
60,000

5,000

5,000

20, 000

5,000

2,000

200

2,000

200
 

 

Estimated Estimated
. Time Cost
Activity Required ($k)
(yr) ('76 $'s)
3. Study of an integrated Canadian-U.S.
Nuclear growth scenario to show maxi-
mum advantage of CANDU-Th system. This
implies an integrated fuel resource base. 2 500
4. Study the economics, fuel utilization, and
fuel management in high-conversion ratio
systems. Consider the trade-offs in
lattice spacing, specific power, and fuel
assembly design etc. as functions of the
probable ranges of ore, reprocessing,
fabrication, and separative work costs. 2 300
5. Investigation of the power cost economics
and fuel utilization implications of
slightly enriched fuel for CANDUs. 1 300
6. Study the advantages of ThC and Th metal in
CANDU-Th reactors to further optimize the
fuel utilization. 1 200
7. Determination of the optimum degree of
symbiosis necessary between CANDU plutonium
producers and CANDU thorium burners. Also,
determination of the value of CANDU-Th with
highly enriched 235U makeup. 2 300
C. Irradiation Program in Support of Research
Programs 7 30,000
D. Design Work in Support of Studies including
Reactor Physics Analyses Associated with a
Fueling Evaluation 3 20,000
$200,000

 

®These costs could be much higher if major redesign studies were

required for licensing purposes.

 
APPENDIX L

SUMMARY OF CALCULATIONS AND CALCULATIONAL METHODS

1. U=-235 — U304 relationship

 

 

kg U _ 3(238) _
kg U30g  2(238) + 8(16) - 8480
tails assay
w7 235U
kg 235y -
kg U = ,00711 - .0010 = .00611 .1
= ,00711 - .0020 = .00511 o2
= ,00711 - .0025 = .00461 _ .25
= .00711 - .0030 = .00411 3

kg 235g _ kg_235U kg U

kg U305 kg U kg Us0g (.00511) (.8480)

 

.00433 (.2% tails)

kg U308 1
= no231
kg 235U  .00433

 

tons U30g 231 (2.2)

 

 

 

 

= = ,254]1
kg 23°U 2000
tails assay (% 23950)
Natural
.1 .2 .25 .3 U
kg 235y
.00518 .00433 .00391 .00349 .00603
kg Us0g
kg U3O8
—_— 193 231 256 287 166
kg 235U
tons Uj30g
_ .212 .2541 .2816 . 3157 .1824
kg 235U ~
2. Average in-core residence time

kg HM | | mwd (th)
MW (e) kg HM |
mwd (th)
(yr) [MW(e)]

 

 

t,. (yo)

ol nrs-wikicol
365 (load factor)

 

3. Total inventoryl

tr + t
I, = \——2)1
S tr Y

 

—_ k 1
IS Efi%gj’total inventory
t, = time in reactor (yrs)
tp = ex-reactor time (yrs)
- kg
Ir MW (o) in reactor

4. U-235 cost?

 

$ _ kg U ) s . kg Swu ) $
kg product kg product/ kg U kg product kg Swu

$

S, _ kg product
kg 235U kg 235U
kg product

Assume: $40/1b U30g =‘%§g%

$75/kg Swu

.2 w % 2350 tails assay

 

IThe Use of Thorium in Nuclear Power Reactors, WASH 1097 (June 1969), p. 22.
2AEC Gaseous Diffusion Plant Operations, ORO-658, Appendix 2.

 
 

- Product
Enrichment
w % 235y kg U/kg Product kg Swu/kg Product $/g Product $/g 233U
711 1.000 0 .104 14.63
2.2 3.914 2.602 ' .602 27.37
2.4 4,305 3.018 .674 28.09
3.0 5.479 4.306 .893 29.76
3.4 6.262 5.191 1.041 30.61
3.6 6.654 5.638 1.115 30.97
93.0 181.605 235.550 36.55 39.30

5. PFuel cycle costs

 

Components:

1.

Inventory costs (capital)
total fissile inventory
fabrication cost of first core

thorium inventory (for thorium cycles)
Shipping

Makeup
fissile

thorium (for thorium cycles)
Reprocessing (for recycle cases only)
Fabrication (refabrication)
Spent fuel storage (for non-recycle cases only)

Heavy water makeup (for HWRs only)
Calculations:

1. Inventory

 

 

 

 

 

 

 

 

[ $ } _ [total kaissile}< $ ) (lO/
MW(e) yr fissile MW(e) kg fissile
[ S ] _ l:kg HM in 1st core] ) 10/ )
MW(e) yr ¢ MW(e) \\kg HM fab/ \\y
ab
[ $ } _ [Eg Th in core]( ‘> (10/
MW(e) yrip MW (e) kg Th/ \
. 825 ( $10 )
Assume: o "Th \1b ThO,
[ $ ] (10+3 mills)
mills _ LMW(e) yr $
kw hr kw hr
MW(e) yr
kw hr =(}<£ (hrs)
MW(e) ;;f‘ (load factor)
$
. mills _MW(e) yr . _
* %w hr 2008 for load section = .8
2. Shipping
. s (8) [ ]
Reprocessing MiW(e) yT kg Mi(e) yr
Fabrication
3. Spent fuel storage
Assume: 230 for CANDU, $100 for LWR, 2200 for HTGR
kg kg kg

 
4. Heavy water makeup

Assume: .35 mill/kw hr (ref. 3)

5. Makeup costs

 

{ $ ] _ [kg fissile makeup} ( $ )
- kg fissil
MW(e) yr fissile MW(e) yr g fissile

 

[y ey, - () (o )

Discussion: In calculating fuel cycle costs in the above manner it is
not necessary to assign a value to bred fuel. It is assumed that this
fuel is recycled to the reactor. Ifi_the case of no recycle a charge is
assigned for storing the spent fuel, with no credit for the fissile
inventory in the stored fuel. Burnup costs are assigned on the basis of
the cost of makeup fuel only. The low makeup costs associated with high
gain converters is balanced by the inventory costs associated with high
specific inventories. This is the fairest, simplest way of comparing
fuel cycles and reactor types with vastly different thermal efficiencies,

conversion ratios, burnups, and processing costs.

 

3Private communication from E. Critoph (AECL) to R. Laney (ANL) dated
19 March 1976.
 

 

 

 
APPENDIX M

QUALITATIVE OVERVIEW OF RECYCLE PROCESS STATUS
FOR VARIOUS REACTOR SYSTEMS

The following tables were derived to show the commonality in fuel re-
cycle process for various reactors and reactor fuels and to provide a
qualitative assessment of the current status of development for processes

associated with these fuel types.

To provide the required perspective the required process steps, includ-
ing irradiation proof-testing of the product, were identified generically
for each of five reactor types and a variety of potential fuels. These
are shown in Table M1 and address both recycled fuel and fresh fuel

since in some instances development is required for that fuel derived
from natural sources. For each of the process steps a current develop-
mental status was defined in terms of the normal development stages

shown schematically in Figure Ml. This status is given by the number

in parenthesis for each generic process label.

Finally in Table M2 the commonality in processes is shown by establishing
the development of recycle capability for current LWR fuels as a base
and show which additional features would require additional develop-
ment to establish the technology for recycle of a new reactor or fuel.
It should be emphasized that this assessment is based on an assumed
sequential development with the base case development and subsequent
developments incorporating both historical data and anticipating future
modifications and additions. Using these same ground rules, a relative
order of magnitude cost projection was made, as shown in Table M2. The
absolute values of these projections are highly uncertain. To establish
an absolute cost projection would require the development and assessment
of a detailed experimental plan. However, the relative numbers given in
Table M2 do show the commonality between systems and how, by generally
small incremental development additions, the number of recycle options
and the ability to choose between alternate fuel resources can be

expanded.
.

 

 

 

 

Table M.l. Overview of Processes for Fuel Recycle with Assessment of Status®
Recycle Fuel Processes Fresh Fuel
Reactor System Recycle Fuel Fresh Fuel
Fuel Head-End Separation Fuel Conversion Refabrication Irradiation Proof Fabrication Irradiation Proof
Testing Teating
LWR uo; Chop—l.each(fi) Purex(“a ADU(S) ’ Pellet Rod Bundle(s) l Provcn(s)
(U=~Pu)0, Chop-Leach(fi) Purex'®) oxalate V) Pellet Rod Bundle'?) proven'?)
(?*3y-Th) 0, Cladding Sepatation(” Thotex(z) plus Purex“”h ADU(S) plus Thermal Penetrat[on(” Pellet Rod Bundle(SA) Required
(2*u-Th)0; Cladding Separationt’ Thorex 2 plus PurexAM oy (D) plus Thermal Penetrationt’ Pellet Rod Bundle'?! Required
(Pu-Th)03z Cladding Separation(l) Thorex(z) plus Purex(&fi’b oxalute(3) plus Thermal Penetratlon(J) Pellec Rod Bundle(ZA) Required
LWBR (2?'y-Th)0, Cladding Separatian(l) Thorex(z) plus Purex(“A)h ADU(SA) plus Thermal Penetratlon(l) Pellet Rod Bundle(z) Requlrcd(])
Pellet Rod Bundles'® proven!®
CANDU U0, Chop-Leach(h) Purex(fl) ADU(S)
(U=Pu)0, Chop-Leach(b) Purex(fl) ADU(S) plus Oxala:e(]) Pellet Rod Bundlc(ZA) Requlred(J)
(233y.Th)o, Cladding Separationcl) Thorex(Z) plus Purex(bh) ADU(S) plus Thermal Penetraticn(l) Pellet Rod Bund]e(SA) Required(a)
(2¥3y.mh)o, Cladding Separatlon(l) Thorex(Z) plus Purex(&h) ADU(Z) plus Thermal Penetratton(]) Pellet Rod Bundle(ZA) ch01TEd(2)
FBR
Core (U-Pu)0a Chop—Leach(‘A) Purex(3A) ADU(S) plus Oxalate(]A) Pellet Rod Bundle(z) Requlred(])
(*?3y-Th)o, Chop-Leach(aA) Thorex(z) plus Pure{(JA) ADU(z) plus Thermal Penetratlon(]) Pellet Rod Bundle(ZA) chulred(J)
(Pu-Th)02 .Chnp-Leach(aA Thorex(z) plus Purex(JA) Oxalete(JA) plus Thermal ?enetrntlun(j) Pellet Rod Bundle(zk) Required
2 u-H) Undefined (1’ Thorex () plus Purex(3A) Reduction-nelt ‘!’ Slug Rod Bundle ! Required
(2Vp-p)c Chop—Lem:hu) Purex(jA) Carbothermic Reductlon(z) Pellet Rod Bund!e(z) Required
Blanket vo: Chop-Leach 4 purex? aou ! Pellet Rod Bundle(™*) provent’)
ThO, Chop-Leach(&) Thorex(z) plus Purex(]h} Thermal Penetration(3) Pellet Rod Bundle(SA) Required
HTGR (?3%y-Th)Cc-0 Burn-Leach(Z) Thorex (2 plus purex (34 Resin Kernel®? Blended Bed-Prismatic Block(Z) Required(&)
(**'u-Th)C-0 Burn-Leach(z) Thorex(?) plus Purex ‘A Resin Kernel ‘2’ Blended Bed-Prismatic(Z) Block Required

 

®Numbers in parenthescs are reference fo the current stage of development (see Figure M1).

The suffix "A" denotes a process that has not been applied to the reactor fuel indicated, but that is readily adapted (at the stage indicated) without significant technological barriers,

 
ORNL-DWG 7618737

COLD PROTOTYPE
LOW RADIATION DEVELOPMENT
LEVEL
LABORATORY ENGINEERING
DEVELOPMENT DEVELOPMENT

 

     
 
 

 
   
  

DEMONSTRATION

   

5

 

COMMERCIALIZATION__O

HOT ENGINEERING
DEVELOPMENT

 

Fig. M.l1l. Development sequence for fuel cycle facilities.
Table M.2. Possible Sequential? Development for Fuel Recycle Capability

 

Add 3 Add 4 Add 5 Add 7 Add 8

Base Add 1 LWR CANDU LWBR Add 6 FBR FBR

Processes LWR CANDU Oxide Oxide oOxide HTICR Oxide Metal
(U-Pu)02 (U-Pu)O0> Th Cycles

Th Cycles Th Cycles Th Cycles Th Cycles Th Cycles

 

Chop-Leach R

Cladding Separation R

Burn-Leach R

Metal-Process R

Purex R
Thorex R

ADU R

Oxalate R

Thermal-Penetration R

Reduction-Melt R
Resin Kernel R

LWR Pellet Rod Bundle R

LWBR Pellet Rod Bundle R

CANDU Pellet Rod Bundle R

FBR Oxide Pellet Rod Bundle _ R

FBR Slug Rod Bundle R
Blended Bed-Prismatic Block R

Recycle Irradiation Proof R R R R R R R

Fresh Fuel Irradiation Proof R R R R R

 

Incremental Development Costb 250 50 125 25 50 250 150 200

Cumulative Development Cost? 250 300 425 450 500 750 900 1100

 

_ qAssumes required processes are developed in the base case or a prior additional step for each additional system, If a
process is not available the needed development is indicated by the letter "R". Also if additional development is needed

it will need to be so designated [Eq. (3%, Th)02]. LWR Fuel Refabrication and bundle assembly is not required if the
(U-Pu)02 LWR fuel is done completely remotely. We have assumed this will be the case in this table.

bFor development to demonstration stage (demonstration plant not included) all cost are in millions of dollars and
exclude the cost of irradiation space for demonstration.

 
LWR{(Th) Fuel Cycle Development

 

R&D work required for the LWR thorium fuel cycle involves developing and
demonstrating fuel recycle technology and fuel irradiation performance.
Developing recycle technology is required to facilitate the implementa-
tion of the thorium cycle in LWRs and to obtain associated improved fuel
utilization.

LWR(Th) fuel recycle technology development could be carried out in an
integrated program with both HWR(Th) and HTGR thorium cycle work since
there are common areas of development. It is estimated that the R&D
costs of LWR thorium fuel recycle would be about $150 million above that
associated with developing fuel recycle technology for the uranium/
plutonium cycle in LWRs. In addition, a demonstration-scale facility
involving both reprocessing and refabrication should be operated, and
the cost of that would be several hundred million dollars above that

associated with industry support.

A fuel irradiation testing and evaluation program would also be required
to qualify thorium-based fuel and recycle fuel, the cost of which would
be about $30 million. An extensive core design effort would also be
needed to determine practical thorium-plus-fissile loading which are
also relatively economical. The cost of such core design and associated

core physics analyses are estimated to be about $15 million.

Overall, close coordination with industry would be required, with most

of the fuel testing done in commercial facilities; where practical, use
should also be made of industrial fabrication and refabrication facili-
ties. Fuel cycle analysis work should also be closely coordinated with

industrial studies.

With regard to the above, past and present development work on the LWR
fuel cycle should be utilized insofar as possible. The large amount of

fuel fabrication effort by industry to date should be at least partially
M-6

applicable; similarly, industrial work on fuel refabrication should be I
utilized. However, it is believed important that the sphere-pac method
be emphasized in the refabrication of fuels. A corresponding fuel test-

ing program would need to be included.

Fuel reprocessing studies can build on technologies previously and
presently being developed; thus, the work being carried out on the Acid-
Thorex Process for thorium fuels will be largely applicable. The shear-
leach process for the head-end processing LWR uranium fuels will provide
very useful information. There are special problems with thorium fuels
(such as those associated with dissolution of thoria), and these will
require specific R&D. Waste disposal treatment studies can build on
those being done for the uranium cycle, taking into consideration any

special requirements introduced by use of thoria.

Insofar as fuel testing is concerned, the irradiation testing performed
under the LWBR program should provide very useful technology information, |I

and such work should form a base for future development and testing.
HTGR Fuel Recycle Development -

The HTGR fuel recycle R&D and demonstration program has been developed

in detail in a National Program Plan document developed jointly by

Oak Ridge National Laboratory, General Atomic Company and Allied Chemical
Corporation under ERDA sponsorship. The R&D effort is estimated to be
about $400 million and the first-phase demonstration plant is estimated

to cost about $600 million.
APPENDIX N

POWER COST AND ORE-UTILIZATION SUMMARY

Summary: Appendices B, C, and D have described metal loadings and makeup
requirements for uranium and thorium fuel cycles in LWRs, HTGRs, and HWRs.
Generally, higher conversion ratios can be achieved with higher initial
loadings and lower burnup. Information on ore utilization has been
presented, but very liftle has been said about cost. Fuel cycle cost
calculations are presented in this Appendix, using the method outlined
in Appendix L. Fixed ore and separative-work costs ($40/1b Uszgg and
$75/SWU respectively) are assumed in these calculations. Variations of
fuel cycle cost with changing ore and SWU costs are discussed elsewhere
in this report. The report confirms that costs associated with high
initial inventories and low burnup outweigh the advantages of better ore
utilization at current ore and SWU costs. It is also shown that the use
of Pu as fissile material is economically superior to U-235, considering
the assumed reprocessing costs. However, since the supply of Pu is
limited, the economics of fuel costs are of littlé value when selecting
reactor systems and fuel cycles to develop.

Relatively low fuel cycle costs (by comparison) were calculated for
several HWR fuel cycles. There is some question in the authors' minds
whether the inventory and makeup requirements reported in the HWR studies
cited in this report are of the same quality as those reported for LWRs
and HTGRs. Higher inventory and makeup requirements would result in
higher fuel cycle costs. In addition, the fabrication, refabrication,
and reprocessing costs used in the HWR fuel cycle cost calculations were
those associated with very large industries, which would be required for
HWRs because of the low exposures achieved in those reactors. Using
costs associated with smaller fuel cycle industries would make the HWR
fuel cycle costs much less attractive.

Considering ali options, it is the authors' opinion that the HTGR
offers the best combination of low fuel cycle cost and good ore utili-
zation. Lowest power generation costs were calculated for the HTGR,
with a conversion ratio of 0.66 and U-233 recycle. Higher ore and SWU

costs would make the higher-conversion-ratio HTGR most attractive.
Fuel Cycle Costs

 

Tables N.1l through N.5 contain fuel cycle cost estimates for the reactors
and thg fuel cycles included in this study. The (a) portions of the tables
summarize the performance data from which the estimates are made, and the
(b) portions contain the actual estimates. Table N.1 covers the PWR, using
various U0,-ThO,, U0p-PuO;, U-Th, and Pu-Th fuel cycles. Table N.2 covers
the PWR with a TRRC (Thorium Replacement Reactor Core), with conversion
ratios of 0.83 to 0.96. A discussion of all PWR options is included in
Appendix B. Table N.3 covers the HTGR, which is discussed in Appendix C.
Table N.4 covers the CANDU using the plutonium cycle, as discussed in
Appendix D. Table N.5 covers the CANDU using the thorium cycle, which is

also covered in Appendix D.

Several aspects of the fuel cycle cost calculations deserve special

mention:

1. LWR fuel cycle cost data presented in Table N.1 are based on
selected studies from Appendix B. There is some variation among studies
(shown in Table B.4) on initial inventories, makeup, and conversion ratio.
There is fairly good agreement between the costs shown in Table N.1l and
costs calculated for similar fuel cycles from other studies, shown in
Appendix B. For example, the UO; with no recycle, described in Table B.2,
has a calculated fuel cycle cost of 6.6 mills/kwhr compared to 6.9 mills/
kwhr in Table N.1. For U0O,-ThO,, the comparison is 9.1 mills/kwhr vs
9.4 mills/kwhr in Table N.1l. The Pu0,-U0, case is an exception. In
Table N.1l, the value is 9.1 mills/kwhr compared with 7.5 mills/kwhr
using data from the BMI study cited in Appendix B. Using data from
Table B.2, the value for PuOy, — natural U0 is 6.3 mills/kwhr. This value
can be explained because of the $20/g assumed for Pu vs $30/g for U-235
(3%Z enriched). Based on these comparisons, it is possible that the value

of 9.1 mills/kwhr, shown in Table N.1, might be high.

 
Table N.1(a).

Parameters for Fuel Cycle Cost Calculation

 

Reactor pwR? PWRa’b
Fuel U0, (No Recycle) U05-Pu02
Conversion Ratio 0.61
Enrichment, % Fissile 3.2 3.2
Thermal Efficiency, % 33 33
Load Factor 0.8 0.8
Burnup (MWd/kg HM) 33 33
Time in Reactor (yr) 3.19 3.2
Ex-Reactor Time (yr) 1.8 2.6
(tr + tp)/tr 1.56 1.81
Inventory [kg/MW(e)]
Reactor/Total 233y
235y 2.74/4.29 2.74/4.96
238U
Fissile Pu
Th
Total 85.6 85.6
Makeup [kg/MW(e) yr]
233y
235y 0.91 0.54
238y
Fissile Pu
Th
Total 26.75 26.75
$/g Fresh 235y 30 30
$/g Makeup Fissile 30 30
30 yr 235y kg/Mi(e)® 30.03 18.94

PURY? b
U0,-ThO,
0.76

4.0

33

0.8

34.5
3.20

2.6

1.81

3.30/5.97
0.25

77.93/141.1

81.5

0.399

25.47
40
40
15.27

PWRb’c
PuO5-ThO5
0.78

4,48

33

0.8

33

3.10

2.6

1.84

3.720/6.73
79.32/144

83.04

0.310

25.9
40
20

PWRa’b

U-Th
0.81
3.44
33
0.8
25.8
3.2
2.6
1.81

3.771/6.836
0.284

105.51/190.97

109.56

0.34

34,24
40
40
13.97

4.41
2.6
1.59

4.398/6.99
114.1/181.4

118.5

0.302

26.87
20
20

 

O U0

External source of Pu assumed.

N

Reactor inventory + 30 years makeup.

Table B.3, Appendix B of this report (GAC entries).
Self generated recycle.
Table B.1l, Appendix B of this report.
Table N.1(b).

Fuel Cycle Cost Calculation

 

Reactor

Fuel

Conversion Ratio

$/kg HM
Shipping
Makeup
$/g 23%
$/g 235U
$/g Pu
$/g Th
Reprocessing
Fabrication
Refabrication
Storage

mills/kW hr
Inventory (fissile)
Inventory (Th)
First Core Fab
Shipping
Makeup
Reprocessing
Fab/Refab

Storage

Total

 

PWR PWR PWR PWR PWR PWR
U0, (No Recycle) UO02-Pu03 U02=ThO> Pu0O2-ThO> U-Th Pu-Th
0.61 0.76 0.78 0.81 0.81
10 10 10 10 10
30 30 40 40
20 20
0.025 0.025 0.025 0.025
292 292 316 300 300
150 150 200 510 200 500
500 570 510 500 500
100
1.836 2.123 3.408 1.921 3.902 1.995
0.050 0.051 0.068 0.0647
0.183 0.184 0.233 0.604 0.313 0.845
0.038 0.038 0.038 0.038 0.049 0.038
3.896 2.590 2.277 0.885 1.941 0.862
1.187 1.133 1.210 1.466 1.147
0.574 1.912 2.212 1.885 2.443 1.912
0.400
6.927 8.034 9.351 6.594 10.182 6.864

 

 
Table N.2(a).

Parameters for Fuel Cycle Cost Calculation

N-5

 

Reactor

Fuel

Conversion Ratio

Enrichment, 7

Thermal Efficiency, %

Load Factor

Burnup (MWd/kg HM)

Time in Reactor (yr)

Ex-reactor Time (yr)

(t, + tp)/tr,

Inventory [kg/MW(e)]
Reactor/Total 233U

235y

238U

Fissile Pu
Th
Total

Makeup [kg/MW(e) yr]
‘ 233y

235U

238U

Fissile Pu
Th
Total
$/g Fresh 23%y
$/g Makeup Fissile
30 yr 235U kg/MW(e)

PWR
TRRC
0.96

0.8
10
0.975
2.6
3.67

2.09/7.67
0.16

84.20/309
86.25

0.13

86.4
88.46
40

40
5.99

PWR
TRRC
0.93

33
0.8
15
1.46
2.6
2.78

2.14/5.95
0.16

84.09/234
86.39

0.158

57.6
59.17
40

40
6.88

PWR
TRRC
0.88

33
0.8
20
1.96
2.6
2.33

2.24/5.22
0.17

84.36/197
86.77

0.180

43.0
43.04
40

40
7.64

PWR
TRRC
0.83

33
0.8
25
2.45
2.6
2.06

2.46/5.07
0.19

86.74/179
89.39

0.224

35.4
36.49
40

40
9.18

 
Table N.2(b).

Fuel Cycle Cost Calculation

 

Reactor

Fuel

Conversion Ratio

$/kg HM
Shipping
Makeup
$/g 233y
$/g 235y
$/g Pu
$/g Th
Reprocessing
Fabrication
Refabrication
Storage

mills/kW hr
Inventory (fissile)
Inventory (Th)
*First Core Fab
Shipping
Makeup
Reprocessing
Fab/Refab

Storage

Total

 

PWR PWR PWR PWR
TRRC TRRC TRRC TRRC
0.96 0.93 0.88 0.83
10 10 10 10

40 40 40 40
300 300 300 300
200 200 200 200
500 500 500 500
4.378 3.396 3.140 3.030
0.110 0.083 0.070 0.064
0.246 0.246 0.246 0.246
0.123 0.082 0.061 0.051
0.742 0.902 1.027 1.278
3.787 2.533 1.840 1.560
6.311 4.220 3.068 2.600
15.697 11.462 9.452 8.829

 

 
Table N.3(a).

N-7

Parameters for Fuel Cycle Cost Calculation

 

Reactor

Fuel

Conversion Ratio
Enrichment, 7

Thermal Efficiency, %
Load Factor

Burnup (MWd/kg HM)
Time in Reactor (yr)
Ex-reactor Time (yr)
(t, + tp)/tp
Inventory [kg/MW(e)]

Reactor/Total 233y
235y

238y

Fissile Pu
Th
Total

Makeup [kg/MW(e)yr]
233U

235U
238U

Fissile Pu
Th
Total
$/g Fresh 235y
$/g Makeup Fissile
30 yr 235U kg/MW(e)

HTGRZ

UC 0 , ThO
X'y

0.66

2

40
0.8
86.4
4.0
1.8
1.45

1.4/2.03
0.1

32.3/46.8
33.8

0.625

8.08
8.45
40

40
20.15

UC_0 _, ThO
Xy

HTGR—lb

2
0.66

40
0.8
86.4
4.0
2.6
1.65

1.4/2.31
0.1

32.3/53.3
33.8

0.324

8.08
8.45
40

40
11.12

uc_ 0 _, ThO
Xy

HTGR-3b

2
0.82

40
0.8
49.7
3.5
2.6
1.74

1.89/3.29
0.11

49.4/86.0
51.4

0.20

14.1
14.69
40

40
7.89

 

No Recycle.

b233U Recycle.
N-8

Table N.3(b). Fuel Cycle Cost Calculation

 

Reactor HTGR? HTGR-1 HTGR-3
Fuel choy’ ThO2 choy’ ThO2 choy’ ThO
Conversion Ratio 0.66 0.66 0.82
$/kg HM
Shipping 60 60 60
Makeup
$/g 233y
$/g 23%U 40 40 40
$/g Pu
$/g Th
Reprocessing 707 707
Fabrication 400 400 400
Refabrication 652b 652b
Storage . 200
mills/kW hr
Inventory (fissile) 1.159 1.318 1.878
Inventory (Th) 0.017 0.019 0.031
First Core Fab 0.193 0.193 0.293
Shipping 0.072 0.072 0.126
Makeup 3.567 1.849 1.142
Reprocessing 0.852 1.482
Fab/Refab 0.482 0.786 1.367
Storage 0.241
Total 5.731 5.089 6.319

 

 

aNo Recycle.
b(0.6) 1030 + 0.4 (400) = 652
Table N.4(a).

Parameters for

Fuel Cycle Cost Calculation

 

Reactor

Fuel

Conversion Ratio

Enrichment, %

Thermal Efficiency, %

Load Factor

Burnup (MWd/kg HM)

Time in Reactor (yr)

Ex-reactor Time (yr)

(£, + tp)/tr

Inventory [kg/MW(e)]
Reactor /Total 233y

235y

238U

(Fissile) Pu
Th
Total

Makeup [kg/MW(e) yr]
233y

235U

238U

(Fissile) Pu
Th
Total
$/g Fresh 235y
$/g Makeup Fissile

CANDU
uo;
0.74
0.711
30
0.8
7.5
1.0
1.8
2.83

0.910/2.58

128

0.910

128
15

20
30

CANDU
Pu02-U02
0.74
1.02

30

0.8

18

2.4

2.6

2.08

0.401/0.834

0.904/1.88

128

0.167

0.376

20
50

CANDU
PuQ2-U02
1.0

1.8

30

0.8

10

1.2

2.6

3.17

2.07/6.56

115

95.8

20
50

CANDU
Pu02-U0,
0.96

1.9

30

0.8

20

2.4

2.6

2.08

2.07/4.31

0.12/0.250

115

0.048

47.9

20
50

CANDU
Pu02-U02
0.93

2.0

30

0.8

25

3.0

2.6

1.87

2.07/3.87

0.23/0.43

115

0.077

38.3

20

CANDU
Pu02-U02
0.90

2.2

30

0.8

33

4.0

2.6

1.65

2.07/3.42

0.46/0.759

115

0.115

28.8

20

CANDU
Pu0,-U02
0.87

2.4

30

0.8

40

4.7

2.6

1.55

2.07/3.21

0.69/1.07

115

0.147

24.5

20
50

CANDU
Pu03-U02
0.85

2.6

30

0.8

44

5.2

2.6

1.50

2.07/3.11

0.92/1.38

115

0.177

22.1

20
30

 
Table N.4(b).

Fuel Cycle Cost Calculation

 

Reactor
Fuel

Conversion Ratio

$/kg HM
Shipping
Makeup
$/g 233y
$/g 235¢
$/g Pu
$/g Th

Reprocessing

Fabrication

Refabrication

Storage
Mills/kW hr

Inventory (fissile)
Inventory (Th)
First Core Fab

Shipping
Makeup

Reprocessing

Fab/Refab

Storage

Total

 

CANDU CANDU CANDU CANDU CANDU CANDU CANDU CANDU
U0, Pu03-U0; Pu05-U0, Pu0,-U02 Pu0,-U04 Pu02-10> Pu0,-U0, Pu05,-00;
0.74 0.74 1.0 0.96 0.93 0.90 C.87 0.85
3 3 3 3 3 3 3 3
15
20 (50)% 20 (50) 20 (50) 20 (50) 20 (50) 20 (50) 20 (50)
147 147 147 147 147 147 147
60 230 230 230 230 230 230 230
230 230 230 230 230 230 230
s0°
0.552 0.775 (1.938) 187 (4.68) 1.30 (3.25) 1.23 (3.075) 1.19 (2.975) 1.22 (3.05) 1.28 (3.20)
0.109 0.42 0.377 0.377 0.377 0.377 0.377 0.377
0.055 0.023 0.041 0.021 0.016 0.012 0.010 0.009
1.945 1.550 (3.875) 0 0.137 (0.343) 0.220 (0.550) 0.328 (0.820) 0.420 (1.050) 0.505 (1.263)
1.118 2.010 1.005 0.803 0.604 0.514 0.464
1.096 1.749 3.144 1.572 1.257 0.945 0.804 0.725
0.913b
4.670 5.635 7.442 4.412 3.903 3.456 3.345 3.360
(9.123) (10.247) (6.568) (6.078) (5.733) (5.806) (6.038)

 

Numbers in parentheses reflect $50/g Pu instead of $20/g Pu.

bStorage costs are based on $50/kg HM.
On this basis, the storage cost would be 0.182 mills/kWhr, and the total fuel cycle cost would be 3.393 mills/kWhr.

other reactor types.

Canadians are presently paying $10/kg HM for storage (personal communication from E. Critoph — AECL).
While the Canadian
value may be more realistic for their purposes, we believe the $50/kg HM value provides a more realistic comparison with costs assumed for

OTI-N

 
Table N.5(a).

Parameters for Fuel Cycle Cost Calculation

N-11

 

Reactor CANDU CANDU CANDU CANDU CANDU
Fuel U02-ThO; U0z-ThO32 U02-Th02 U02z-ThO; U02-ThO;
Conversion Ratio 0.90 0.87 0.82 1.0 0.93
Enrichment, % 2.0 2.2 2.7 1.72 1.88
Thermal Efficiency, % 30 30 30 30 30
Load Factor 0.8 0.8 0.8 0.8 0.8
Burnup (MWd/kg HM) 15 27 44 8.5 27
Time in Reactor (yr) 1.52 2.74 4.47 1.29 4.11
Ex-reactor Time (yr) 2.6 2.6 2.6 2.6 2.6
(tP + tp)/tr 2.71 1.95 1.58 3.02 1.63
Inventory [kg/MW(e)]
Reactor/Total 233y
23%0  1.74/4.72 1.91/3.72 2.35/3.71  2.24/6.76  2.44/3.98
238y 0.13 0.14 0.18 0.17 0.18
Fissile Pu
Th 85.13/358 84.95/166 84.47/133 127.59/385 127.38/208
Total 87 87 87 130 130
Makeup [kg/MW(e) yr]
233y,
235y 0.109 0.140 0.183 0 0.076
238y
Fissile Pu
Th 56 31 18 98 31
Total 57.24 31.75 19.46 100.78 31.63
$/g Fresh 23°y 40 40 40 40 40
$/g Makeup Fissile 40 40 40 40 40
30 yr 2°°U kg/Mi(e) 5.01 6.11 7.84 2.24 4.52

 
Reactor

Fuel

Conversion Ratio

$/kg HM
Shipping
Makeup
s/g 23
$/g 235U
$/g Pu
$/g Th
Reprocessing
Fabrication
Refabrication
Storage

mills/kW hr
Inventory (fissile)
Inventory (Th)
First Core Fab
Shipping
Makeup
Reprocessing
Fab/Refab

Storage

Total

Table N.5(b).

CANDU

N-12

Fuel Cycle Cost Calculation

 

CANDU CANDU CANDU CANDU
U0,>-ThO, UO2~ThO> UQ2-ThO» U02-ThO» U02-ThO,
0.90 0.87 0.82 1.0 0.93
3 3 3 3 3
40 40 40 40 40
0.025 0.025 0.025 0.025 0.025
157 157 157 157 157
60 60 60 60 60
293 293 293 293 293
2.694 2.118 2.118 3.858 2.271
0.127 0.059 0.047 0.137 0.074
0.074 0.074 0.074 0.111 0.111
0.024 0.014 0.008 0.043 0.014
0.622 0.799 1.044 0 0C.434
1.277 0.694 0.403 2.263 0.717
2.383 1.338 0.794 4.223 1.338
7.201 5.096 4.588 10.635 4.959

 

 
N-13

2. The lowest fuel cycle costs are calculated for HWRs with conver-
sion ratios around 0.85. The lowest costs were calculated using the uranium
cycle with plutonium topping and assuming $20/g for Pu (LWR-discharge Pu).
If CANDU-discharge Pu were used, the costs were much higher. The thorium
cycle also looked attractive in the HWRs, although the fuel cycle costs
were calculated on the basis of a large recycle industry, which would be
necessary because of the low exposure achieved in HWRs. Fuel cycle costs
would be much higher initially until a large recycle industry is established.
In this regard, the HTGR looks more attractive because of the relatively low
fuel cycle costs calculated on the basis of a small recycle industry. 1In
any case, heavy water makeup costs push the HWRs to higher total power costs
than HTGRs, as will be discussed later in this Appendix. It is concluded,
therefore, that the HTGR with U-233 recycle offers the best combination of

resource utilization and low power cost.

3. Fuel cycle costs for the Pu-Th cycle are considerably lower than
those for the U-Th cycle, with comparable conversion ratios. This differ-
ential is due to the costs assumed for makeup uranium and plutonium. Makeup
uranium (93% enriched) was assumed to cost about $40/g, which reflects an
ore cost of $40/1b of U30g and a separative-work cost of $75/SWU. Makeup
plutonium was assumed to cost $20/g, which is the approximate cost assumed

for reprocessing LWR fuel. This will be discussed further.

4. Fuel cycle costs for nonrecycle in LWRs on the uranium cycle and
in HWRs on the natural uranium cycle were higher than the costs for recycle
cases. This differential is due to the high cost of reprocessing and re-
fabrication relative to the cost of fresh fissile fuel. If the cost of

recycle fissile material is computed by Eq. (N1)

($/kg) = ($/kg) + ($/kg)

fissile repro refab (slkg)fab ? (N1)

and the reprocessing, refabrication, and fabrication costs described in

Appendices H and I are assumed valid, the comparisons shown in Table N.6
Table N.6. Comparison of Fresh and Recycle Fissile Material Costs

 

 

 

$/kg $/kg $/kg kg Fissile $/¢ Fissile? $/g Reczcleb
Reprocessing Refabrication  Fabrication kg HM Discharge & $/g Fresh
LWR 292 500 150 0.015 43 1.5
CANDU (5 tonne/day reprocessing 271 310 80 0.003 167 11.1
2 tonne/day fabrication
2 tonne/day refabrication)
CANDU [50 GW(e) economy] 147 230 60 0.003 106 7.1
HTGR (5 tonne/day reprocessing 707 1030 400 0.03 44.6 1.1
2 tonne/day fabrication
2 tonne/day refabrication)
HTGR [50 GW(e) economy] 1233 1576 612 0.03 73.2 1.8
a - -
(3/8) fissile (slkg)repro + (slkg)refab (slkg)fab'

bs/g) fresh

Values based on $40/1b U30g and $75/SWU separative work.

Y1-N

 

 
N-15

can be made. Notice the penalty associated with using Pu from CANDUs. A
high concentration of fissile material in the discharge is needed to make

reprocessing profitable.

5. The costs for metal fuel reprocessing, fabrication, and refabri-
cation were assumed to be the same as oxide fuel. This assumption probably
penalizes metal fuels, since significant cost savings are envisioned if
metal fuel and cladding could be coextruded. No meaningful studies of this
fabrication route have been made, and metal fuels are of little interest

for water reactors at present.

Other Costs

A summary of all power costs is tabulated in Table N.7. Only fuel cycles
requiring uranium makeup are included. Fuel cycles requiring plutonium
makeup from other reactors are generally less expensive, as noted earlier;
however, the calculation of ore utilization is complicated by the need to
consider the amount of ore used to generate the plutonium. As noted
earlier, the HTGR (U-233 recycle) has the lowest fuel cycle costs, followed
by the CANDU-Th with a conversion ratio of 0.82. The following additional

comments on the information in Table N.7 are of interest:

1. Capital costs are based on an estimate of $900/kW(e) for LWRs! and
the capital cost ratios given in the BMI study.2

2. Heavy water costs are from ref. 2.

3. The ore utilization capabilities of the various reactors and fuel
cycles are based on the performance data tabulated in Tables N.1
through N.5. There is some variation in this data and that pre-
sented in other studies. Table N.8 contains a comparison of the
ORNL (this study), ANL (ref. 2), and BMI (ref N1) studies for
cases that appear in all three studies. This comparison shows that
the ORNL estimate for LWR fuel utilization is high but in good
agreement with at least one other study for the CANDU and HTGR.
Table N.7. Cost Summary for Thorium-Uranium Fuel Cycle Alternatives

 

mills/kW hr

 

 

e
Reactor roel o Reactor® b [ oC b Cycle Dzod Povar Bgegzizzz‘zfie Numgfigpoziege:;tors ‘éearsf
Capital 2 Maké&up Cost kg 235u/Mw 3.5 x 10% tons U304 upply
PWR vo,? 0.61 19.3 2 6.9 28.2 30.03 459 23
UOp-ThO;  0.76 19.3 2 9.4 30.7 15.25 903 45
U-Th 0.81 19.3 2 10.2 29.7 13.97 986 49
U-Th 0.88 19.3 2 9.5 30.8 7.64 1803 90
HTGR uco-Tho,?  0.66 19.5 2 5.7 27.2 20.15 684 34
HTGR-1 UCO-ThO;  0.66 19.5 2 5.1 26.6 11.12 1239 62
HTGR-3 UCO-ThO,  0.82 19.5 2 6.3 27.8 7.89 1746 87
HWR ThOp-U0,  0.90 21.4 2 2.6 7.2 0.4 33.6 5.01 2749 137
0.87 21.4 2. 2.6 5.1 0.4 31.5 6.11 2254 113
0.82 21.4 2 2.6 4.6 0.4 31.0 7.84 1757 88
1.0 21.4 2 2.6 10.3 0.4 36.7 2.24 6149 307
0.93 21.4 2 2.6 5.0 0.4 31.4 4.52 3047 152
CANDU vo,9 0.74 21.4 2 2.6 4.7 0.4 31.1 28.21 682 34

9T-N

 

aCapital costs based on $900/kW(e) for LWRs, and a capital cost ratio of other reactors to LWRs of 1.11 for CANDUs
and 1.01 for HTGRs. A fixed charge rate of 15%/yr is assumed.

o

Operation and Maintenance costs assumed the same for all reactors and fuel cycles.

O

D,0 costs based on $120/kg D0 and a requirement of 1 MT/MW initial inventory of heavy water in CANDUs.
D0 makeup costs from private communication E. Critoph (AECL) to R. Laney (ANL) dated March 19, 1976.

® R,

The number of reactors supported by 3.5 x 10% short toms U30g assumes 0,2% tails concentration (except for
CANDU where 0% tails concentration is assumed) or 0.2541 tons ore/kg 23°U (0.182 for CANDU). The reactors assumed
to be 1000 MW(e).

f&he Years Supply column assumes a linear growth rate of 20 MW(e)/yr for nuclear power. The entry in this
column is then the number of 1000 MW(e) reactors divided by 20 MW(e)/yr. No more reactors could be built after
the year specified without exceeding the 3.5 million tons of ore. Reactors already on line would operate to
the end of the 30 yr economic lifetime.

gNo Recycle.

 

 
N-17

Table N.8. Comparison with Other Studies

 

 

ANL ‘ - BMI ORNL
Conversion Ratios
LWR 0.61 0.55 0.61
CANDU 0.74 0.71 0.74
HTGR © 0.66 0.65 0.66
30 Year Ore Requirementsa
LWR
No Recycle 6.8 6.2 7.6
U Recycle
U, Pu Recycle 4.8
CANDU
No Recycle 5.2 3.6 5.1
Pu Recycle 3.1 3.0
HIGR
No Recycle 4.8 4.5 5.1
U-233 Recycle 3.0 2.8

 

%0re utilization is short toms Us0g/MW(e).

requirements include the first core loading plus 30 years of makeup.

Thirty year ore
N-18
REFERENCES FOR APPENDIX N

1. A Brief Survey of Congiderations Involved in Introducing CANDU
Reactors into the U.S., unpublished Argonne National Laboratory,
Argonne, I1l., report.

2. W. M. Pardue et al., "A Comparison of Advanced Reactor Potentials,"
paper presented at ASME/ANS International Conference on Advanced
Nuclear Energy Systems, March 14-17, 1976, Pittsburgh, Pa,

 
Appendix O

IRRADIATION PERFORMANCE OF THORIUM=-CONTAINING FUELS

Summary: The irradiation performance of thoria, thoria-urania, and thorium-
uranium metal fuels is reviewed. Thoria and thoria—-urania fuels appear to
be well behaved. These fuels perform at least as well as U0, under the same
reactor conditions. Qualification and specification development for thoria-
urania should be possible in a reasonably short time, and at modeSt expense.
The data base on thorium carbide and (Th,U)C, comes mostly from the HIGR
fuel development program. Continued testing supported by that program have
shown oxides or oxide-carbide fuels to be superior in performance to the
carbides. Therefore ThC, and (Th,U)C, are of little further interest in
HTGRs. Thoria-plutonia fuels have not been extensively tested. There

does not appear to be any reason why performance of this system should

not be as favorable as the thoria-urania system, but because of the lack

of data, the qualification and specification development program would
require more time and be more costly than a similar program for thoria-

urania.

Thorium and thorium—uranium metal fuels look attractive for FBR-application.
The irradiation stability of compounds containing less than 20 wZ U irradi-
ated below 650°C looks good. Thorium-plutonium compounds do not appear useful
as nuclear fuels because of the formation of low melting Pu-rich phases.
Thorium-uranium metal fuels for LWRs are not seriously considered because of
the potential for metal-water reactions. While the thorium-water reaction
proceeds at a rate in thorium which is two orders of magnitude lower than in
uranium at the water temperatures of interest, the development program
required to qualify and specify such a fuel does not appear justified on

the basis of the small improvement in conversion ratio that might be |

achieved.
Thorium and Thorium-Uranium Oxides (ThO, and (U,Th)0,)

 

Thorium oxide has been studied more extensively than any other thorium
compound. A number of irradiation experiments involving ThO, are reported

in reference 1, including:
1. dense pellets with 6.36 w% U0, in the Borax IV BWR blanket;

2. the first cores of the Indian Point PWR and Elk River BWR, which also
used pressed and sintered pellets of ThO,~U0,;

3. PyC coated ThO, microspheres have been extensively tested with the

support of the HTGR fuel development program;

4. coated particles of (U,Th)0, have been extensively tested as potential

HTGR fuels.

Thoria-urania fuels have been considered for several reactor concepts,

1 an HWR design using concen-

including the Spectral-Shift Converter Reactor,
tric fuel tubes filled with vibratory compacted ThO,-U0, mixtures,? and the

Heavy Water Organic Cooled Reactor (HWOCR).3

A detailed summary gf the irradiation behavior of ThO, and (Th,U)O, has been
published by Olsen. In this work, the irradiation behavior of ThO; and
(Th,U)0, in three different forms were compared. The forms were (1) vibratory
compacted sol-gel powder, (2) arc-fused (Th,U)O, rods, and (3) rods containing

pressed and sintered pellets.

The conclusion reached by Olsen et al.* is that all three forms of thoria-
urania fuel performed well at burnups up to 80 MWd/kg HM. There was no

evidence of breakaway swelling or sudden increases in fission gas release.
The average linear heat rates for these fuel rods were between 300 and 350

w/em (9.8 to 11.5 kw-ft).

 
Thoria-Plutonia Fuels (Th,Pu)0O,

 

Very little work has been done with this fuel. One (Th,Pu)0, fuel rod was
included in the work described by Olsen et al.,"* but examination of this rod
was incomplete at the time reference (4) was written. Preliminary examina-
tion of this fuel, which had been irradiated to a burnup of 29 MWd/kg HM at
an average linear heat rate of 245 w/cm (8 kw-ft), showed a microstructure
similar to (Th,U)O, irradiated under the same conditions. Clearly, the
deficiency of information about the performance of thoria-plutonia fuels
must be addressed if large scale use of the thorium fuel cycle in LWRs and

CANDUs is to be seriously considered.

Thorium Carbide and Thorium-Uranium Carbide (ThC, and (U,Th)Cy)

 

Most of the irradiation experience on this system has been accumulated by
General Atomic Company (and other HTGR proponents) on coated-particle systems
for the HTGR. The Fort St. Vrain Reactor (FSVR) core is a (U,Th)Cp fissile
and ThC, fertile combination. Subsequent developments in HIGR fuel technology
have shown oxide and mixed oxide—carbide fuels to be superior in irradiation
performance to the carbides. Therefore, the carbide system will probably not
be considered further in HTGRs. It could probably be considered for other

reactors such as the organic-~cooled HWRs.

Metal Fuels

In Appendices B and E the performance advantages of thorium-metal fuel for
LWRs and FBRs, respectively, were discussed. Higher conversion ratios are
possible with metallic fuels because of the higher concentration of fissile
atoms and the lack of oxygen atoms to absorb neutrons. It is anticipated
that some savings in fuel fabrication cost could be realized with thorium-

uranium metal alloys if coextrusion techniques can be developed.

The use of metallic fuels in LWRs has never been seriously considered
because of dimensional instability in metallic uranium and problems
caused by metal-water reactions. Thorium metal behaves better than

uranium on both counts.
0-4

The dimensional stability of metallic thorium under irradiation at temper-
atures below 600°C is well known,® and the corrosion rate of bare thorium
metal is two orders of magnitude lower than the rate for uranium in water

at 300°C, as shown in Figs. 0.1 and 0.2 from reference 6.

No significant change in the corrosion behavior of thorium metal has been

observed when alloyed up to 6 w% uranium.®

Another major difference in the behavior of thorium-based fuels compared
with uranium-based fuels is the mechanism of failure. In uranium-fueled
rods, corrosion products block the cladding breach, permitting only leakage
of water into the failed rod. No release of fuel to the coolant occurs
until the blister formed ruptures and enlarges the breach. Fuel then exits
to the cladding on a massive scale. With thorium—based fuels, both hydrogen
and finely divided oxide are discharged from the breach continually. The
initial discharge of fuel to the coolant signals the location of the failure
before large amounts of fuel have entered the coolant. Shutdown of the
reactor can be accomplished without major inconvenience because of the

early warning and the relatively low rate of fuel corrosion.®

Even given the advantages of thorium metal fuel over uranium metal in water
reactors, the development program required to qualify metal fuel for this
application is probably not justified on the basis of the small gains in
conversion ratio which can be achieved. However, in a FBR with sodium or
gas coolant, the water-metal reaction problem does not exist, and the gains
in breeding ratio achieved with metal fuel appear well worth the development

cost.

Figure 0.3 summarizes the irradiation performance of a number of thorium-
uranium metal alloys, irradiated at temperatures up to 1000°C and burnups
of up to 10% FIMA. The swelling rate remains constant at about 2% volume
increase per % FIMA, up to about 500°C. At higher temperatures, higher

swelling rates are observed, and a strong temperature sensitivity exists.

Volume increases measured in this work® were linear with burnup and

 
0-5

ORNL-DWG 76-17700

 

TEMPERATURE (°C)
5 315260 200180 100
O r—r—T171 I

WEIGHT LOSS (mg/cm®/hr)

 

   

UNALLOYED
URANIUM

  
  

 

 

10 —
10° |- —
o't | |
160 200 240 280
1000
T°K
Fig. 0.1. Water corrosion of uranium vs water temperature.

Source:

G. B. Zorzoli, "An Evaluation of a Near-Breeder, Low Cost,

LWR Concept," Energia Nucleare 19(3) (March 1972).
ORNL-DWG 76-17698
TEMPERATURE (°C)
102 315260 200480 100
T 11 [
\
\
\
o~ \\
= 10'L -
> \
£ \
Q \
g \
< 5 ‘5 UNALLOYED
@ 107 THORIUM  —
S
|_
I
= ot
i e -
= 10
fo°L 1 |
1.60 200 240 280
1000
T°K
Fig. 0.2. Water corrosion of thorium vs water temperature.

Source: G. B. Zorzoli, "An Evaluation of a Near-Breeder, Low Cost,

0-6

 

 

 

 

 

LWR Concept," Energia Nucleare 19(3) (March 1972).
14

12

10
wija

s|S5 8
3%
>
®

T o

4

2

0o

ORNL-DWG 7647697

 

 

 

 

 

 

 

A Th-5.

 

A Th-{

 

 

LEGEND

e THORIUM
© Th-O4{ w/o U
X Th-44w/o U

5w/o U

OTh-{Owo U

Sw/o U

8 Th-20w/0 U
V Th-25w/o U
¥ Th-31{ w/o U

 

 

 

 

W

 

 

 

 

 

 

 

 

 

 

 

:f A
|

 

o L

v
—

S

 

4

 

 

 

Qo
O e

}It,

 

 

 

 

 

 

 

 

 

 

 

0

100

200 300

400 500

600 700

800 900

MAXIMUM CENTRAL IRRADI_ATION TEMPERATURE (°C)

Fig.
of thorium and thorium-uranium alloys.

0.3.

Source:

Effects of Irradiation on Thorium and Thorium Alloys, ANL-5674
(Apr. 1, 1963).

1000

Effect of irradiation temperature on the swelling rate
J. H. Kittel, et al.,

1100
0-8
independent of uranium content up to 20 w%Z U. Specimens containing more l
than 25 wZ U became warped and distorted, a condition not noted in speci-
mens containing less than 20 w%Z U. Similar results were reported by

workers at Battelle’ and Atomics International®.

Thorium-plutonium metal alloys have been rejected as candidate fuels.? The
low melting point of plutonium compared with thorium and uranium (640°C for
Pu, 1750°C for Th, 1132°C for U) is a potential source of problems for binary
or ternary systems containing Pu. Compositions tending to form Pu-rich
phases would have to be avoided because such phases would be expected to

have melting points near that of Pu metal. Thorium-plutonium alloys tend

to form Pu-rich phases as U-233 is bred in.l0
10.

0-9

REFERENCES FOR APPENDIX O

The Use of Thorium in Nuclear Power Reactors, WASH-1097 (June 1969)
Appendix B.

M. W. Rosenthal et al., A Comparative Evaluation of Advanced Converters,
ORNL-3686 (January 1965).

WASH-1083.

A. R. Olsen et al., "Irradiation Behavior of Thorium-Uranium Alloys and
Compounds,'" IAEA Technical Report Series No. 52, Utilization of Thorium
in Power Reactors (1966), also available as ORNL/TM-1142 (June 1965).

J. H. Kittel et al., Effects of Irradiation on Thorium and Thorium
Alloys, ANL-5674 (April 1963).

G. B. Zorzoli, "An Evaluation of a Near-Breeder, Low Cost, LWR Concept,"

Energia Nucleare (19), No. 3 (March 1972).

J. E. Gates et al., The Examination and Evaluation of Irradiated
Thoriwm-11 w% Uranium Specimens, BMI-1334 (April 1959).

B. R. Hayward et al., Radiation Behavior of Metallic Fuels for Sodium
Graphite Reactors, NAA-SR-3411 (August 1959).

R. P. Hammond et al., The Unclad-Metal Breeder Reactor (UMBR) for
Degalting or Power, ORNL-4202 (January 1969).

L. R. Kelman et al., "Status of Metallic Plutonium Power-Breeder Fuels,"
Plutonium 1965, Proceedings of the Third International Conference on
Plutonium, London, 1965, Chapman and Hall, London, 1967 (Chapter 22).
 

o e - 3 T T T e s . I

 
 

 

 

 

 

 

 

 
)

APPENDIX P
COMMENTS ON FISSILE AVAILABILITY FOR FBR ECONOMY

Summary: A simple nuclear growth model is considered for comparing the
fissile inventories available for Fast Breeder Reactors (FBRs). A
growth rate of 20 GW(e) per year (LWRs without recycle) over a 30-year
period, followed by a 30-year period of constant power at 600 MW(e),
will consume all of the currently estimated uranium resources. If
plutonium is used in FBRs, and the nuclear growth rate is sustained at
20 GW(e) per year, only about 40% of the energy production in LWRs over
the first 30 years can be diverted to non-plutonium producing fuel
cycles. Use of an HTGR-233U FBR system results in greater fissile inventory
available for FBR startup, and in a lower mined ore requirement. The
above assumes that HTGRs are available on the same basis as LWRs; similar

results would apply if HWRs were used instead of HTGRs.

The current policy, with respect to development of nuclear energy from
fission reactors, is to build around breeder reactors which can supply
excess fissile material as well as produce power. New reactors can be
built from the supply of excess fissile material without resorting to
mining of additional uranium. Because of the importance of the breeder
reactor and its early introduction, the growth of the breeder economy
must not be constrained by lack of adequate fissile inventory. Simply
put, there must be enough plutonium discharged from LWRs to sustain the
growth of LMFBRs until the LMFBRs are producing enough plutonium to
sustain their own growth. This report has discussed the virtues of
thorium fuel cycles in converter reactors. Improved resource (U30g)
utilization can be achieved by employing thorium fuel cycles in HTGRs,
CANDUs, and even LWRs as compared to the uranium cycle in LWRs. In some
cases superior economics can also be achieved. However, it must be
recognized that whenever natural 235y is employed in a thorium fuel

cycle, the stockpile of plutonium available for FBRs suffers.

The fissile availability problem is placed in perspective by the following

example. Assume a nuclear growth rate of 20 GW(e) per year. For the
P-2

first 30 years, LWRs are built and the discharged plutonium stored.
Beginning in the 31st year, LMFBRs (based on advanced oxide fuel) are
built at the rate of 20 GW(e) per year; further, LWRs are retired at the
rate of 20 GW(e) per year, and replaced with new LWRs using the inventory
of the retired plants (plus a small amount of fissile makeup.) This case
is shown graphically in Fig. P.l. The reactor fissile requirements are
given in Table P.l for the LWR, advanced oxide LMFBR, and several other

reactors considered in a comparison study that will be discussed.

The case shown in Fig. P.1l for LWRs followed by LMFBRs is designated
Case A, and the plutonium inventory as a function of time is shown in
Fig. P.2. 1If no 235y is used after year 30, and all fissile material
for startup of LMFBRs and refueling of LWRs must come from the approxi-
mately 1500 tonnes of plutonium stockpiled in year 30, growth in nuclear
capacity can only continue for 5 more years, as shown by the solid
portion of the Case A curves in Fig. P.2. If 235U is used to refuel the
600 LWRs on line, and the plutonium stockpile is used only for startup
of the LMFBRs, then the plutonium stockpile is reduced to about 600
tonnes in year 55, when the discharge of excess plutonjum from LMFBRs
begins to exceed the plutonium required for startup of new reactors.
This situation is shown by the dashed curve for Case A in Fig. P.2.
Unfortunately, this case required more than 3.5 million tons of Uj30g;
about 4.0 million tons are used by the year 60. Even so, it is clear
than less than 40%Z of the energy production during the first 30 years of
Case A can be accomplished using thorium fuel cycles, if the plutonium
availability constraint is not to be violated. Plutonium availability
considerations also affect the use of plutonium-thorium cycles, or

plutonium recycle in LWRs.

Another case is presented for comparison. This is Case B, where the
converter reactors built during the first 30 years are HTGRs, with a
conversion ratio of 0.82; after this time, retired HTGRs are replaced.
Uranium-233 is bred and stored for use in 233y fueled FBRs, beginning in
year 31. The characteristics of 233y FBRs presented in Appendix E have

been assumed and summarized in Table P.l. Because of the superior

 
P-3

ORNL-DWG 76-18734
1200

 

__1000

 

)

 

BREEDER

 

800

 

400

 

 

 

CONVERTER

INSTALLED NUCLEAR CAPACITY [GW(e
N
3

 

3

 

 

 

 

 

 

 

 

0 10 20 30 40 50 60
YEARS

Fig., P.l. Nuclear Growth Model for Fissile Inventory Study.
ORNL-DWG 76-18733

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

3000
\~ //
case "8"[ |\\\, ,
2500 x N~
\
\ “-METAL FUELED
\\ 233 FBR; MINED
\ |FUEL FOR CONVER-
5000 \| TER MAKEUP
3 OXIDE FUELED FBR; \
£ MINED FUEL FOR
g CONVERTER MAKEUP | \
: -/ \
& \
= 1500 |———1 X
L NO 2% miNED |/ \ \
> AFTER YEAR 30 \ N
\ N
W \ ™~
5 \\ e
¥ 1000
T
\
\
L
// CASE nAn \ \.____..__.,/
500 / :
0
0 10 20 20 40 50 60

YEARS

Fig. P.2, Fissile Inventory Versus Time for U-Pu System (Case "A')
and U-Th System (Case "B"). )

 
Table P.1. Reactor Characteristics for Fissile Inventory Study

 

Fast Breeder Reactors

 

 

e Hror %gvanceda Th Metal® Th Metal®  Th M0x°  Th MOX®
Nauéogigg Na Cooled He Cooled Na Cooled He Cooled
Reactor Requirements
Reactor Inventory
[kg/MW(e)]
233y 1.572 2.143 1.879 2.910
233y 1.967 1.890 0.064
Fissile Pu 2.080
Ex-Reactor Inventory
[kg/MW (e)]
233y 1.572 2.143 1.879 2.910
233y 1.107 1.399 0.064
Fissile Pu 2.080
Annual Loading
[kg/MW(e) yr]-
233y 0.524 0.714 0.626 0.970
233y 0.754 0.540 0.030
Fissile Pu 0.785
Annual Discharge
[kg/MW(e) yr]
233y 0.310 0.733 0.938 0.757 1.14
233y 0.222 0.162 0.025 ‘
Fissile Pu 0.164 0.958
Conversion Ration 0.60 0.82 1.25 1.26 1.29 1.16 1.21
Doubling Time (yr) 24.0 15.1 19.1 28.8 34.3

 

4. M. Pardue et al, "A Comparison of Advanced Reactor Potentials' presented at the ASME/ANS International
Conference on Advanced Nuclear Energy Systems, March 14-17, 1976, Pittsburgh, Pa.

bUnpublished data of M. Z. Nagel et al (General Atomic Company) "Reactor Strategy Studies" 19 January 1976.

cAppendix E of this report.

G-d
P-6

conversion characteristics of the HTGR, the available fissile inventory
(233y in this case) for FBRs in year 30 is approximately twice that in
the LWR-LMFBR case. Even using the poorest performing FBR discussed in
Appendix E, the fissile inventory in year 60 is over 1000 tonnes of
233y, as shown in the dashed portion of Case B plotted in Fig. P.2. .
Using more optimistic FBR performance data (metal fuel), the 233y

inventory by year 60 has returned to the level in year 30, as shown by

the broken Case B curve in Fig. P.2. Even if no 235y is used after year

30, there is an adequate 233y inventory to fuel both FBRs and HTGRs for

about 13 years (compared with 5 years for the LWR-LMFBR case). As shown

in Fig. P.3, only 2.6 millions toms of U30g have been consumed in Case

B, compared to 4 million tons in Case A.

It is clear from this simple example that the HTGR-233U fueled FBR com-

bination has superior capabilities with respect to resource utilization

and fissile inventory for FBRs, if they were available on the same

basis. The economics of the HTGR relative to the LWR are also favorable, .
as shown in Appendix N. Conversion ratios of 0.9 and greater appear .
achievable with the HTGR utilizing current fuel technology, with no

apparent sacrifice in reactor safety.

Little advantage can be taken of the Case B findings because neither the
HTIGR nor the 233(U-fueled FBR are being pursued seriously at this time.
However, even with delayed introduction of HTGRs some of the above
benefits can be obtained. Based on recent studies summarized in Appendix .
F of this report, 233y use in LMFBRs (with 2380) appears as attractive

as does plutonium.
P-7

ORNL-DWG 76-18732

35 /

CASE "A"

30 //r

2.5 /
// CASE "B’ /

[/

 

 

 

 

 

 

 

 

Uz0g USED (short tons x10~6)

ASSUMES MINED ORE
USED FOR CONVERTER

 

 

 

 

 

 

 

 

 

 

MAKEUP
1.0 /
0.5 /
0 :
0 10 20 30 40 50 60
YEARS

Fig. P.3. U30g Requirements for U-Pu System (Case "A") and U-Th
System (Case "B").
ey ey hfl_flqxflxfi‘.flft

FEE T - - ¢ e TR e e e e e e S A Tt ans e o -

 

" .-.grr??i “‘fl
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o T T

 
APPENDIX Q

CONSIDERATIONS REGARDING BREAK-EVEN BREEDERS

Summary. A simple example case is described wherebfi an HTGR thermal
breeder (HTGRB) system is compared with several conventional converter
reactor systems over a 60-year period. Reactors are built at a rate of
20 GW(e)/year for the first 30 years and then allowed to '"coast down"
over the second 30-year period. Compared are the time-dependent power
production capabilities and uranium ore requirements. It is shown that
the thermal breeder systems give a long-term power production capability
which do not require additional uranium ore input after the initijal 233y
inventories are produced in pre-breeders. However, the early ore require-
ments are considerably higher for the thermal breeder systems than for
conventional converter systems over the time span considered. Ore re-
quirements are 64% higher than for an HTGR (CR = 0.82), and fuel cycle
costs are two to three times higher than for conventional HIGRs and

LWRs. Similar results should also apply to the use of HWR(Th)s. It is
concluded that while thermal breeder systems provide long-term power
production capability at a fixed level, there is no flexibility for

growth provided by these sytems for a given U30g resource. Thus, they
essentially represent a contingency position relative to a very long-

term delay in FBR commercialization.

Break-even breeder systems are possible with LWRs, HWRs, and HTGRs.
Such systems require production of 233y inventories in pre-breeder
reactors; the thermal breeders then operate on a 233y-Th fuel cycle.
The advantages and disadvantages of such break-even systems can be seen

with the following example.

A nuclear growth rate of 20 GW(e)/year was assumed; pre-breeders were
built first to provide 233y inventories for the break-even breeders,
which were built as soon as inventories were available. Power growth
remained constant at 20 GW(e)/year for the first 30 years. During the

next 30 years, pre-breeders were retired at the end of their economic
Q-2

lifetime (30 years of operation). Inventories of 233U from retired
reactors and from annual discharges of operating reactors were utilized
to build additional break-even breeders during the years 31-60., All
other fissile material discharged from the pre-breeders was recycled to

the pre-breeders.

Considered in the example was a break-even HTGR system using 233y
inventories produced in a lower-conversion-ratio HTGR. The mass require-
ments for the various reactors considered in the example are given in
Table Q.1. The installed capacity vs time relationship is shown in Fig.
Q.1 for the HTGR breeder (HTBR-B) system, for a light-water reactor With
no recycle, and for an HTGR with a conversion ratio of 0.82 and 233U
recycle., The ore requirements are shown in Fig. Q.2 for the breeder
system, the LWR with no recycle, the HTGR (CR = 0.82), and several other

systems which are included for comparison.

For the HTGR breeder system, a total of 334 pre-breeder HTGRs (HTGR-PB)
were built, producing enough 233y for 498 HTGR breeders (HTGR-B).

Fuel cycle cost estimates are presented in Tables Q.2 and Q.3. In these
estimates, it was assumed that 23%U was purchased for pre-breeder oper-
ation, but 233U was made available to the breeders at no cost. On this
basis the HTGR-PB and HTBR-B had fuel cycle costs of about 12 mills/kWhr.
From these results it appears that the economic advantages in long-term,
self-sustaining power generation possible with break-even breeders are
outweighed by the high cost of operation and the large uranium ore
requirements of the pre-breeders. Any economic advantages associated
with the breeder operation have difficulty in being 'visible" after

applying a reasonable discount factor.

Three criteria have been identified for an effective thermal reactor
system. They are:
1. Make effective use of uranium resources in the period
prior to large-scale fast breeder reactor (FBR)

introduction.
Table Q.1. Reactor Characteristics

for Break-Even Breeder Study

 

 

Reactor Requirements LWRb HTGRC HTGR
Reactor Inventory [kg/MW(e)]
233y .960%
235y 1.967 3.700
Fissile Pu
Ex-Reactor Inventory [kg/MW(e)]a
233y .874
235y 1.107 2.410
Fissile Pu
Annual Loading [kg/MW(e)yr]
233y . .490
235y . 754 .925
Fissile Pu
Annual Discharge {[kg/MW(e)yr] |
235y 222 .258
Fissile Pu .164
tr (Reactor Time) (yrs) 3.0 4.0 .0
tP (Ex-Reactor Time) (yrs) 1.8 2.6
(t, + L)/t 1.60 1.65 .65
Conversion Ratio .60 .74 .01

 

a . .
Ex-reactor inventory = reactor inventory [(t, + tp)/tr]

Source: W. M. Pardue et al, A Comparison of Advanced Reactor Potentials, Presented at the
ASME/ANS International Conference on Advanced Nuclear Energy Systems, March 14-17, 1976,

Pittsburg, Pa.

cSource: Private Communication from R. K. Lane (GA) to F. J. Homan (ORNL), June 14, 1976.

Source: Letter from R. F. Turner (GA) to E. DeLaney (ERDA), May 28, 1976.

£-0
ORNL-DWG 76-18735

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

600
\
\ HTGR-PB; HTGR-B
500 \ £ >

5 \\
= \\
E QOO \
2 \ HTGR (CR=0.82)
< \
° \
qu 300 \
w
- \
> \
= \
o 200 \
-
3 LWR-—/’\\
. (no recycle) \
= 100 3

\

\

0
o - 20 40 60 80 100 120 140 160
YEARS
Fig. Q.1. Installed capacity is time for LWBR, LWR, and HTGR
comparison.

-0

 
ORNL-DWG 76-18736

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

40 [
~
/
X LWR (U, Pu recycle) /]
3 |+ LWR (U recycle) /
| 0 HTGR (CR=066 233U recycle) | /
A& HTGR (CR=066 no recycle) /
v CANDU (CR=0.74 no recycle) X
20 * HTGR (CR=082 233U recycle) 2
O CANDU-Th(CR=082) " / 1
/
& /
O /
=< 25
X /
o /
tfl : LVVR]—;—‘——tnv
— no recycle —
— / ’
§ 20 /
Q
O fiJ
& HTGR PB
o HTGR-B
o 1.5
2 —
©
10
——HTGR CR=082
05
O
0 10 20 30 40 50 60
YEARS
Fig. Q.2. Ore requirements to sustain a 20 Gw(e)/year nuclear

growth rate for 30 years.
Q-6

Table Q.2. Parameters for Fuel Cycle Cost Calculation

 

 

Reactor HTGR-PB” HTGR-B”
Fuel
Conversion Ratio 0.74 >1.0
Enrichment (%) 93
Thermal Efficiency (%) 40 40
Load Factor 0.8 0.8
Burnup (MWd/kg HM) 30.4 18.8
Time in Reactor (yr) 4 4
Ex-Reactor Time (yr) 2.6 2.6
(tr + 1:p)/tr 1.65 1.65
Inventory [kg/MW(e)]
Reactor/Total 233U 5.960/9.834
235y 3.70/6.11
238y
Fissile Pu
Th 92.5/152.6 149.0/245.9
Total 96.2 155.0
Makeup [kg/MW(e) yr]
233y 0
235y 0.668 0
238y

Fissile Pu

Th 23.1 37.3

Total 24,1 38.8
$/g Fresh 235y 40
$/g Makeup Fissile 40

 

 

YHGTR-PB data from private communication from R. K. Lane (GA) to F. J.
Homan (ORNL) June 14, 1976.

bHTGRrB data from letter to E. Delaney (ERDA) from R. F. Turner (GA),

dated May 28, 1976.
 

 

 

Table Q.3. Fuel Cycle Cost Calculation
Reactor HTGR—-PB HTGR-B
Fuel 2359C0-Tho, 2333C0-ThO5
Conversion Ratio 0.74 >1.0
$/kg HM |
Shipping 60 60
Makeug
2 3U
235U 40 _
Pu
Th 0.025 0.025
Reprocessing 707 707
Fabrication 400 1030
Refabrication 450 1030
Storage - -
Mills/kWhr
Inventory (fissile) 3.487 -
Inventory (Th) 0.054 0.088
1st Core Fab 0.549 2.278
Shipping 0.138 0.332
Makeup uranium 3.813 -
Thorium 0.082 0.133
Reprocessing 2.431 3.914
Fab/ReFab 1.548 5.703
Storage
Total 12.102 12,448

 

L-D
Q-8

2. Provide sufficient fissile inventory for commerciali-
zation of FBRs.

3. Economic power production.
Thermal breeder systems do not meet these criteria effectively. Thermal
breeder systems provide the means for long-term power production at a
constant level, with no additional uranium ore requirements. However,
the high ore requirements needed to provide the 233y inventories of the
breeder systems occur during a time of projected ore shortages, and no
fissile inventories are built for FBRs other than those associated with
the break-even breeders themselves. It is our conclusion that only if
FBRs are delayed until about 2100 should break-even breeders of the type
studied here be developed. Further, the power costs of such systems
will be prohibitively high under the conditions of introduction assumed
here. A less costly approach would be to gradually increase the conversion
ratio with time, although such an approach leads to a lower power level

achieved by the break-even breeders.

@.

 
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