ORNL-TM-9780-V2.txt

  

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NATIONAL
LABORATORY

 

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MARIETT.

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2 ORNL/TM-9780/V2

Nuclear Power Options
Viability Study

Volume I,
Reactor Concepts, Descriptions.,
and Assessments

D. B. Trauger
D. White
. T. Bell

S. Booth
|. Bowers

C. Cleveland
. G. Delene
ri Gat
. C. Hampson

J
J
R
H
J
J
U
D

Jenkins

L. Moses
E. Pasqua
L. Phung
. Spiewak
E

T
D
P
D
|

R. E. Taylor

APPLIED TECHNOLOGY

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Releasad far anntuncement
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R34T, DOE
 

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P.O. Box 62, Oak Ridge, Tennessee 37830

 

 

 

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ORNL/TM-9780/V2
Distribution Category UC-79T
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NUCLEAR POWER OPTIONS VIABILITY STUDY

VOLUME 1II,
REACTOR CONCEPTS, DESCRIPTIONS, AND ASSESSMENTS -

"y

ights. Refer-
, recom-

ts use would not infringe privately owned ri

D. B. Trauger, Editor
Je Do White

J. T. Bell

R. S. Booth

H. I. Bowers

Je Ce Cleveland
J+ G. Delene

U. Gat

D. C. Hampson
T. Jenkinsl

D. L. Moses

P. E. Pasqua2
D. L. Phung3

I. Spiewak"

R. E. Taylor1

y legal liability or responsi-
tion, apparatus, product, or
y thereof. The views

ponsored by an agency of the United States
ent nor any agency thereof, nor any of their
» process, or service by trade name, trademark

necessarily constitute or imply its endorsement

d States Government or any agenc
herein do not necessarily state or reflect those of the

DISCLAIMER
agency thereof.

express or implied, or assumes an

completeness, or usefulness of any informa

lTennessee Valley Authority
2The University of Tennessee
3professional Analysis, Inc.
“Consultant

, or represents that i

ence herein to any specific commercial product

manufacturer, or otherwise does not

Date Published -~ September 1986

Government. Neither the United States Governm

This report was prepared as an account of work s
employees, makes any warranty,

bility for the accuracy,

process disclosed

mendation, or favoring by the Unite

and opinions of authors expressed

United States Government or any

Prepared for the
Office of the Assistant Secretary for Nuclear Energy
U.S. Department of Energy

Prepared by the
OAK RIDGE NATIONAL LABORATORY
Oak Ridge, Tennessee 37831

operated by !4
MARTIN MARIETTA ENERGY SYSTEMS, INC. ffij \ £ }gp
F

for the

U.S. DEPARTMENT OF ENERGY
under Contract No. DE-AC05-840R21400 Released for annoum:fme?l
In ATE. Bigtribution !imned_ E;
pariicipants in he Lmir R
pragran. Others request 110
asot, DOE, kfi}
L
PREFACE

The Nuclear Power Options Viability Study (NPOVS) was initiated at
the beginning of calendar year 1984, The objective of NPOVS was to ex-—
plore the viabilities of several nuclear options for this country for
electric power generation after the year 2000. The study emphasized
technical issues but also considered institutional problems. Innovative
reactor concepts were identified which may be marketable at the time
when studies show that the demand for new electrical energy capacity
will dincrease significantly. These concepts were considered with em-—
phasis on cost, safety features, operability, and regulation as well as
research needs. The study 1s reported in four volumes. Volume I is an
executive summary. This report, Volume II, provides descriptions and
assessments with respect to criteria established in the study of
potential nuclear power plants which could be deployed early in the next
century. Volume III, Nuclear Discipline Topics, provides supporting
analyses; and Volume IV is a bibliography containing approximately 550
entries, A detailed outline covering all four volumes is given in
Appendix A,

The study was initiated by Oak Ridge National Laboratory (ORNL),
which, recognizing the need for a broad base of knowledge and exper-—
ience, engaged the Tennessee Valley Authority (TVA) and The University
of Tennessee to participate as partners. TVA concentrated its efforts
on evaluation of the concepts and on licensing. The University of
Tennessee assisted in the evaluation of construction costs and public
opinion issues. Both institutions contributed extensively to the
evaluation of issues and in review of reports. In conducting the study,
the authors extensively contacted segments of the nuclear industry for
current information concerning the concepts studied and for other valued
assistance,

Many of the problems encountered by the nuclear industry are insti-
tutional in nature and are related to the way the utility companies,
designers, constructors, and regulators are organized and function.
Although this study attempted to identify those institutional factors,
it has not addressed them in all aspects. It was observed that the in-
stitutional problems derive in some measure from technical aspects,
which, in turn, originate at least in part from the large size, com-
plexity, and exacting requirements for existing nuclear plants.
Emphasis in the study was placed on technical aspects that have poten-
tial merit and on improved design concepts that may help or have promise
of helping to alleviate institutional problems. TInstitutional factors
related to market acceptance have also been surveyed and studied. Con-
sideration of additional institutional factors is thought to be desir-
able, perhaps necessary, but is beyond the scope of this study.

The study emphasized criteria by which nuclear power reactors can
be judged and which are thought to be appropriate, at least in part, for

iii
judging future commercial viability. Other design or operational needs
that are important but are more difficult to quantify are presented as
either essential or desirable characteristics. Several innovative
reactor concepts are described and evaluated with respect to these
measures. Related and generic information on construction, economics,
regulation, safety and economic risk, waste transportation and disposal,
and market acceptance which supplements the evaluation is included in
Volume III. '

This study differs in several respects from other studies concern-
ing the future of nuclear power in the United States. The first is the
time frame of interest. The NPOVS effort was focused on a time frame a
little later than most studies, the years 2000 through 2010. For the
near term, existing Light-Water Reactor (LWR) designs, or evolutionary
modifications to them, would be the most likely nuclear choices if there
is a sufficient demand for increased electrical generating capacity.
Projections by the electric industry indicate that new base load capac-
ity will be needed before the year 2000. Therefore, it is probable that
decisions to order baseload capacity will be made by 2000-2010 and,
furthermore, that the reactor concepts discussed in this report have the
potential for competing with existing IWR designs or coal-fired plants
at that time. For the more distant future, nuclear plant concepts
incorporating more innovative, if not revolutionary, features could be
the best choices.

A second aspect making this study different is the level of tech-
nical detail in the evaluation of the specific designs. Significantly
more design information was generated by all the nuclear designers in-
volved with innovative concepts in the last three years, and much of
this information was made available to NPOVS. Recognition has been
given to the special features of each concept and thus to the role that
each may achieve in a mature nuclear economy.

Systematic development of the information presented in this report
was completed in September 1985. Delays in funding and review have pre-
vented timely publication. An attempt has been made to include new
information where substantial changes 1in programs or designs have
occurred, but it has not been possible to bring the report fully up to
date. Subsequent developments and events, particularly the Chernobyl
accident, may alter some of the findings.

iv
ABSTRACT

SUMMARY OF FINDINGS

1.

INTRODUCTION
BACKGROUND

1.1

CONTENTS

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1.2 REPORT ORGANIZATION <accas

1.3 REFERENCES FOR CHAPTER 1

GROUND RULES AND CRITERIA .eese
GROUND RULES AND THEIR SIGNIFICANCE ccoccesccccscsssnas

2.1

2.2 CRITERIA

2.3 REFERENCES FOR CHAPTER 2

CONCEPT EVALUATIONS

3.1

3.2

3.3

2.2.1

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Listing of Criteria, Essential and Related
Desirable Characteristics, and Discussion
of Their ApplicationsS ececesssccccescsccsovencscs

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5 85 ¢ 5 08080 H B e e OB S BB PR eS e

CONCEPT SELECTION AND CLASSIFICATION cscocecosscscnsscs

CONCEPT EVALUATION METHODS AND LIMITATIONS «ceevecccocss

LIGHT WATER REACTORS (LWRs)

3.3‘1

3.3.2

PIUS S8 PO 3 5P 0SS 000 SR LN PSR E eSS eSS SRS

3.3.1.1

Description

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3.3.1.2 (Claims, Advantages, and

Disadvantages Evaluated

Against Criteria, Essential

and Desirable CharacteristicS ..eecesces
R&D Needs and Open Questions

3.3.1.3

Evaluated

Small Advanced BWR

3.3.2.1
3.3.2.2

3.3.2.3

Description

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Page

ix

3-17

3-21
CONTENTS (CONTINUED)

3.4 LIQUID METAL COOLED REACTORS (IMRS) sevevcacssessansnss
3.4.1 Introduction
3.4.2 Design Options, Challenges, and Tradeoffs ......
3.4.3 Design Descriptions seeeeseeccccssccsscssrscccsns

3.4.3.1 The Large Scale Prototype

Breeder (LSPB) cceccsesccccssscosscssos
3.4.3.2 Sodium Advanced Fast Reactor

(SAFR) cecsessessossssssassscsasassnnsnse
3.4.3.3 Power Reactor—-Inherently Safe

Module (PRISM) cecessssocosssscecanssocsse

3.4.4

26 8 05 5 0 0 P OO PP B e S SE S S eEeEEeS

Advantages and Disadvantages of

the IMR Concepts with Regard to
the NPOVS Criteria and Essential
CharacteristicS eeceessvsssocsssscascssscsssnanse
3.4.4.1 General OVerviewS .seesecessccssssssscos
3.4.4.2 Advantages of the IMR ConceptsS sesseses
3.4.4.3 Disadvantages of the IMR

CoNnceptsS seesscsssssssnesccscsccansccnes
3.4.5 Research and Development Needs for the

IMR Concepts

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3.4¢5.1 IntroductiOn eeceescsccscscssssscsscsssns
3.4.5.2 Design-Specific R&D Requirements seeees
3.4.5.3 General R&D Goals for the U.S.

National IMR Program escescevecvecsccccsce

3.5 HIGH TEMPERATURE REACTORS (HTRS) ccecoesscccsscocssassse

3.5.1
3.5.2

345.3

Design DesScriptions ceecececssccccrosssssccrcsss
Claims, Advantages, and Disadvantages

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and Desirable Characteristics escscesscsccncaccss

Modular HTR Research and Development

Needs Evaluated eeseeessscsscsscsssvccascssssssns
3.5.3.1 Base Technology .csececssssscsscscacons
3.5.3.2 Applied Technology ceeccesesscscosccses
3.5.3.3 Design and Economic Studies eececesccse

3.6 REFERENCES FOR CHAPTERB * S 08 080 OB E B OB eSO NESE eSS

4. ACKNOWLEDGMENTS

APPENDIX A:

APPENDIX B:

APPENDIX C:

B 0O E PP P eSO RSSO PSSR S OS S SAEESEEEeTS

BASIC OUTLINE FOR NUCLEAR POWER OPTIONS
VIABILITY STUDY FINAL REPORT «see¢sevsacscccccscccccs

THE OUTLOOK FOR ELECTRICITY SUPPLY
AND DEMAND ® B P F O 8PS NSNS PSS eSS NS EeNE S

DISCUSSION OF CONCEPTS NOT INCLUDED

FOR ASSESSMENT

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vi

Page
3-22
3-22
3-22
3-23
3-24
3-29

3-32

3-35
3-35
3-35

3-39
3~-43
3-43
3-43
3-43
3-45
3-45
3-54
3-58
3-59
3-59
3-60

3-61
APPENDIX D:

APPENDIX E:

APPENDIX F:

APPENDIX G:

CONTENTS

R&D GOALS AND SPECIFIC REQUIREMENTS
FOR LIQUID METAL REACTOR (LMR) CONCEPTS «ceeesssess

LIQUID METAL REACTOR (LMR) FUEL
REPROCESSING-REFABRICATION EVALUATION esoscvesccscs

860 MW(e) LARGE HIGH TEMPERATURE GAS-
COOLED REACTOR (HTGR) a0 00O ¢ OO SIS E SO Es P EE S

EVALUATION OF CLAIMS FOR THE MODULAR HIGH

TEMPERATURE REACTOR (HTR)

vii

Page
ABSTRACT

The Nuclear Power Options Viability Study (NPOVS) is reporting here

on the description and assessment of several selected innovative reactor
designs in accordance with criteria established in the study. These
criteria are as follows:

1.

The calculated risk to the public due to accidents is less than or
equal to the calculated risk associated with the best modern Light-
Water Reactors (LWRs).

The probability of events leading to loss of investment is less than
or equal to 10™% per year (based on plant cost).

The economic performance of the nuclear plant is at least equivalent
to that for coal-fired plants. (Financial goals for the utility are
met, and busbar costs are acceptable to the public utility
commissions.)

The design of each plant is complete enough for analysis to show
that the probability of significant cost/schedule overruns is
acceptably low.

Official approval of a plant design must be given by the U.S.
Nuclear Regulatory Commission (NRC) to assure the investor and the
public of a high probability that the plant will be licensed on a
timely basis if constructed in accordance with the approved design.

For a new concept to become attractive in the marketplace, demon-
stration of its readiness to be designed, built, and licensed and
begin operations on time and at projected cost is necessary.

The design should include only those nuclear technologies for which
the prospective owner/operator has demonstrated competence or can

acquire competent managers and operators.

The criteria are supplemented by essential characteristics that

both amplify the criteria and suggest additional qualities that have a
bearing on viability.

In selecting the concepts to be studied, the following three ground

rules were used:

1.

The nuclear plant design option should be developed sufficiently
that an order could be placed in the time period 2000 through 2010.

The design option should be economically competitive with environ-
mentally acceptable coal-fired plants.

The design option should possess a high degree of passive safety to
protect the public health and property and the owner's investment.

ix
This study led to the choice of the following concepts:
l. Light Water Reactors (LWRs)

® PIUS (Process Inherent Ultimate Safety) — Promoted by ASEA-ATOM
of Sweden

® Small BWR (Boiling Water Reactor) — Promoted by General Electric
(GE)

2. Liquid Metal Reactors (LMRs)

® PRISM (Power Reactor Intrinsically Safe Module) — The advanced
concept supported by DOE

@ SAFR (Sodium Advanced Fast Reactor) — The Rockwell International
(RI) advanced concept supported by DOE

® LSPB (Large—-Scale Prototype Breeder) — The Electric Power Re-
search Institute—Consolidated Management Office (EPRI-CoMO) con-
cept supported by EPRI and DOE

3. High Temperature Reactor (HTR)

® Side—by—-Side Modular — The core and steam generator in separate
steel vessels in a side-by-side configuration. The concept is
supported by DOE and promoted by Gas—Cooled Reactor Associates
(GCRA) and industrial firms.

These concepts are judged to be potentially available in the chosen
time period, are estimated by their promoters to be economically compe-
titive with coal-fired power plants, and have varying degrees of passive
safety attributes. In all cases, the designs are too preliminary for a
complete and definitive assessment, but each is believed to have poten-
tial for a significant future role. The Advanced Pressurized-Water Re-
actor (APWR), the Advanced Boiling-Water Reactor (ABWR), and the 1large
HTR are recognized as viable systems which could meet electric power
generating needs prior to or following the year 2000, They are not in-
cluded in this study except for reference because they do not fully meet
the third ground rule and because they already have been the subject of
extensive study by industrye.
SUMMARY OF FINDINGS

The criteria and the essential and desirable characteristics
described in Chapter 2 are presented as useful guides to the selection
and evaluation of current and future nuclear reactor concepts. The
criteria were developed early in the study and have been subjected to
extensive review and refinement. The criteria serve as guides for the
evaluations of concepts presented in this report.

Most advanced reactor concepts are smaller than present LWRs;
therefore, they suffer the disadvantage, whether real or perceived,
associated with economy of scale. This disadvantage is claimed to be
of fset or compensated for in varying degrees through an improved match
with load growth, reduction in capital risk, increased shop fabrication,
shorter construction time, increased standardization, design simplifica-
tion, and simpler construction management requirements. Licensing also
may be simplified. A substantial problem in achieving these compensa-
tions derives from the need for a large front—-end investment for certain
of these features. Automated shop fabrication, in particular, may re-
quire a substantial backlog of orders to be economically feasible. Nu-
clear plant standardization is widely viewed as an important goal for
viability.

To be concise in this summary, we have assumed that the reader is
familiar with the concepts. If not, the concept descriptions should be
read first. The claims, advantages, disadvantages, and important devel-
opment needs will be summarized in the order in which the concepts ap-
pear in the report.

It must be noted that each of these concepts is currently in design
development. The descriptions and assessments of this study reflect the
status for each as of September 1985 except that some updating has been
done where information was readily available. The reader must recognize
that further development is expected to change design features and thus
to affect the conclusions from future evaluations.

All of the concepts for the study appear to be potentially viable,
but the available information has been insufficient to assess fully
their economic competitiveness. Findings for the concepts are summar-
ized here.

PIUS

The basic design premise of this concept is to achieve a very high
degree of passive safety with respect to equipment failure, operator er-
ror, and external threats. The large pool of borated water, which can
enter the core without mechanical, electrical, or operator action, is to
provide both shutdown and seven days of passive cooling for decay heat
removal. These claims appear to be justified, although questions remain
concerning the safety of fuel and equipment handling operations within
the pool. The availability of water for introduction to the pool after
seven days 1is site dependent but potentially viable. Overall, the

xi
concept appears to be licensable without major redesign, assuming that
the NRC will accept a reactor with no control rods. The features that
promote safety also appear applicable to protection of the investment.
ASEA-ATOM claims that the plant can be economically competitive with
coal-fired plants. This may depend on further evaluation of the identi-
fied problems that follow.

The steam generator located inside the Prestressed Concrete Pres-
sure Vessel (PCPV) is of a difficult design with respect to mainte-
nance. This and the problems of handling fuel and equipment deep (30 m)
in the pool require careful design and detailed assessment and are con-
sidered a disadvantage with respect to potential availability of the
plant. Management of refueling appears difficult for the three-core
design if the refueling sequence becomes out of phase.

Development and testing needs include further demonstration of
fluid interface stability, extensive study and testing of steam gener-
ator modules, thorough testing of underwater fuel and equipment handling
systems, steam flow and pressurizer stability for the multimodular de-
sign, thermal insulation development and testing, and demonstration of
the PCPV design, particularly for the top closure.

Small Advanced BWR

 

This reactor obviously derives advantage from its many operating
gsimilarities to existing BWRs., The gravity—-fed Emergency Core Cooling
System (ECCS) appears well conceived and adequate to provide shutdown
cooling for up to three days, although its reliability is dependent on a
relatively untried fail-open valve. A reliable site-dependent supply of
additional cooling water would be required beyond the three days. Oper-
ator training should be straightforward since the basic operation is
similar to that of existing BWR plants. A first-of-a-kind safety test
and demonstration whereby the plant would later be used for power pro-
duction may be practical and adequate for this concept.

Cost competitiveness is difficult to assess at this early stage of
design development. Liceunsing requirements, although not thought to be
particularly difficult, have not yet been addressed completely. Model
testing for the gravity—-drain ECCS, steam injector testing, and thermal
hydraulic and seismic analyses must be thorough. The depressurization
valve also requires design, development, and testing.

LMR Concegts

The LSPB is an evolution of previous Liquid Metal Fast Breeder
Reactor (LMFBR) designs. It includes several innovations to reduce cap-—
ital costs and to enhance safety. The latter include diverse and inde-
pendent reactor shutdown and dedicated decay heat removal systems.
Although this design offers lower costs, these features are yet to be
evaluated fully with respect to construction methods and

Xii
licensability. The LSPB appears attractive in offering economy of scale
and increased passive safety.

Since the PRISM and SAFR concepts are under development in a DOE
program and are at a preliminary stage, they were assessed primarily in
common. LMR concepts benefit from the inherent IMR features of 1low
system pressure and high thermal conductivity of the 1liquid metal
(sodium) coolant. These features permit the design of primary contain-
ment pool concepts with reliable passive natural convection decay heat
removal to atmospheric air. The smaller modular concepts provide the
potential for testing of the core stability for conditions that might
result from reactivity increases or from 1loss of flow of primary
sodium. In this respect, smaller reactors generally have an advantage
over the larger IMRs. Although the hypothetical core disruptive acci-
dent (HCDA) is claimed to be incredible by the proponents, the case for
disregarding this accident is yet to be approved by the NRC. If the
Clinch River Breeder Reactor (CRBR) can be taken as a precedent, then it
would be reasonable to expect that a HCDA would be a beyond design basis
event (BDBE). Acknowledging passive accommodation of HCDAs could sig-
nificantly reduce design complexity, facilitate licensing, and improve
public acceptance. Reliable control rod shutdown systems, feedback
response from temperature increases, and the resulting thermal expansion
are important safety features. Although containments are provided,
careful design and perhaps testing will be required to ensure that air
oxidation of the sodium cannot occur. Where primary boundaries and con-
tainments are in close proximity, they must be well protected from ex-
ternal threats that could breach both enclosures.

An important advantage of the IMR is the extensive operating ex-
perience available from the Fast Flux Test Facility (FFTF), the Experi-
mental Breeder Reactor-II (EBR-II) and the European and Japanese
plants. However, little of this experience is in the U.S. utility base,
and the loss of the Clinch River Breeder Reactor (CRBR) also slowed the
pace of development of IMRs in this country. Although it appears
possible to design, construct, and operate a demonstration plant and to
reach commercialization within the 2000—2010 time scale, it would
require an early dedication to the task.

The availability of related experience, the simplicity of the pro-
posed designs, and the passive features mentioned above strongly suggest
that the licensing of IMRs should be achievable. An obvious long-term
advantage of the LMR is its potential for breeding and thereby creating
an essentially unlimited extension of uranium fuel resources. Although
this is not an immediate objective of the current program, it should not
be overlooked.

Historically, IMR concepts have had higher capital cost factors
than LWRs. This cost experience is manifested in European as well as
U.S. designs. Although the current concepts address this issue, it is
as yet not adequately resolved. The long life core designs represent
one approach to overall cost reduction. Another disadvantage is the
requirement for enriched uranium or plutonium as the starting fuel. The

xiii
former is essentially assured by present U.S. enrichment programs since
the required production capacity exists. The latter is at an early
stage of development for acceptable fuel reprocessing plants in this
country, although considerable experience exists abroad and in military
facilities. Other countries offer the prospect for purchase of
plutonium, but this is unlikely to be an acceptable continuing source.
Once~-through cycles are not adequate for long-term nuclear energy via-
bility; therefore, reprocessing remains an important objective for
future IMR concepts. The potential use of integral fast reactors which
would be collocated with the supporting fuel reprocessing and refabrica-
tion facility faces significant institutional problems for operation.

Development needs include advanced core design and approved neutron
counting systems, improved shielding, and self-actuated shutdown sys-—
tems. Testing of heat removal systems is an obvious requirement. De-
pending upon the choice of initial fuel, the fuel cycle requirements may
be extensive. The use of metal fuels, which offer some safety and oper-
ational advantages, requires an extensive fuel development and testing
program. Concepts under consideration should benefit substantially from
demonstration reactor testing; however, we caution against overly opti-
mistic expectations from this approach. Indeed, valuable experience can
be gained, and analytical techniques can be tested. However, many dis-
turbances (such as the effects of severe seismic events, external inter-
ference with air cooling, and sabotage) cannot be tested and would
require analysis for evaluation. Such analysis would include probabil-
istic risk assessments and simulations wusing models verified against
data from smaller scale experiments. A more desirable and convincing
approach with respect to utility acceptance may be to construct and
operate a demonstration or prototype plant based on an adequate program
of analysis, component development, and testing and design.

Modular HTR Side-by-Side Concepts

 

A high degree of public protection 1is achieved through the
avoidance of fuel damage by virtue of a capability for extended after-
heat removal through the vessel wall by convection, conduction, and
thermal radiation without operator, mechanical, or electrical interven-
tion. This advantage is made possible by the very high temperature
capability of the fuel, including retention of fission products and the
slow thermal response of the core, which eliminates the need for a fast-
acting shutdown system. The same protection applies to the probability
of events leading to a loss of investment. The "low-enriched"” fuel is
an advantage in proliferation resistance but requires enrichment beyond
that for a conventional IWR. The potential for producing high-temper-
ature process heat is a long-term advantage.

Disadvantages include a potentially high overnight capital cost.
Most, if not all, HTR designs have projected capital costs higher than
those of large IWR reactors. This must be overcome by high availability
and high capacity factors, shop fabrication, reduced costs for the
balance of plant, and low fuel cycle costs.

xiv
A disincentive for the application of an HTR in the United States
is the poor performance to date of the Fort St. Vrain reactor. However,
the difficulties with this reactor are unique to its equipment and are
not common with the new concept. In many ways, the MHTR design can
benefit from the lessons learned at Fort St. Vrain. Also, the per-
formance of the Peach Bottom Unit 1 reactor in the United States and
that of the Arbeitsgemeinschaft Versuchs-Reaktor (AVR) in Germany has
been satisfactory. Since the Thorium High Temperature Reactor (THTR-
300) in Germany is the latest HTR to start operation, its operating
performance is important to watch.

The small base of operating experience in the United States sug-
gests an important need for a successful MHTR demonstration plant.
Because of the ruggedness and resiliency of the fuel and the inherently
slow time response for thermal excursions, it seems possible that a
demonstration plant could be used extensively for experimentation in
safety and operability and could later serve as a first-of-a-kind power
plant. However, since the design experience is also limited, it is
quite possible that a standardized plant design might benefit from the
experience of the demonstration.

Development needs include the determination of fission product re-
tention by the fuel coatings, graphite, and metal surfaces under hypo-
thetical extreme accident conditions. Although acceptable graphite is
available for fabrication of fuel elements and reflector blocks, statis-
tical data are needed to better understand its widely variable proper-
ties as a basis for improvement of the basic structure and to obtain
longer lifetimes. Structural materials development needs relate to
obtaining physical property data to satisfy code requirements, including
the effects of neutron irradiation on physical properties. Another
important area of development is testing of key components under actual
conditions. This testing is especially important in the case of the
circulator and refueling system to assure high availability. Materials
corrosion testing for various projected impurity levels in the helium
would also be desirable but probably is not a requirement.

xv
1. INTRODUCTION

1.1 BACKGROUND

The Nuclear Power Options Viability Study (NPOVS) was begun at Oak
Ridge National Laboratory (ORNL) in January 1984 to assess selected nu-
clear power options with respect to viability and to identify new direc-
tions for industry, regulation, and research. For the first five
months, the study was funded through the ORNL Director's discretionary
fund. Since June of 1984, the U.S. Department of Energy (DOE) has
funded the program directly. The Tennessee Valley Authority (TVA) and
The University of Tennessee (UT) have been working with ORNL in this
study; TVA has used its own funds, while UT has been funded by ORNL
through subcontracts.

The information on which this report is based has been gathered
from the open literature, as well as from reactor design organizations,
vendors, research and development (R&D) institutions, utility companies
and the DOE. Proprietary information or other information received with
restrictions has been considered in the assessment but is not displayed
per se.

In a recent study by the Atomic Industrial Forum, Inc. (AIF), the
following statement is made in the Executive Summary: “Nuclear power
cannot at this time be considered a viable option on which to base new
electric generating capacity in the U.S."! The study points out that
only half as much nuclear capacity will be constructed by 1986 as com-
pared to that which was projected only ten years ago. Since 1972, 108
plants have been cancelled; no new orders have been placed since 1978.
It is significant to note that there were no new large plants of any

type ordered in 1984. Others have addressed projected needs for
electric energy and indicated a rqu}ged increase in baseload capacity
by the year 2000 or soon thereafter. (See also Appendix B).

Several recent studies have considered the issue of nuclear power
viability.7‘9 Most of these either (1) concern themselves with
evaluations of large Light-Water Reactors (LWRs) or (2) assume that
development of a new brand of passively safe reactor designs would lead
to a second wave of nuclear reactor orders. The arguments are centered
around the assumptions that either (1) LWR evolutions could provide by
far the most cost-effective way of meeting the nation's needs or (2) in-
herently safe designs could lead to decreases in licensing time and
costs, be made simpler and cheaper than existing IWRs, and be more
acceptable to the public and the investor.

This study has emphasized technical detail in the evaluation of the
specific designs. Institutional factors are recognized as very impor-
tant, even to the extent of overshadowing technical issues, and are
therefore included in the criteria chosen for evaluating the concepts.
However, the principal thrust of this series of reports is on technical
issues that have merit in their own right and particularly on those that
may alleviate institutional problems. For example, added passive safety
may simplify regulation. Several institutional factors are considered
1-2

generically for the reactor concepts in chapters of Volume III. Signif-
icant new design concepts have been generated in recent years by nuclear
designers involved with innovative approaches. This information consti-
tutes a substantial portion of the subjects considered in this volume.
Attention 1is given in the study to safety and reliability, cost,
licensing, and development needs as well as to the special features of
each concept.

The NPOVS program has proceeded in the following steps: (1) a
literature search and development of a bibliography; (2) development of
criteria for evaluation of nuclear plant designs and plans; (3) evalua-
tion of selected design concepts using these criteria as a guide; and
(4) recommendations for areas of R&D to reduce uncertainties 1in the
viabilities of options. The approach used in evaluation was to compile
detailed information on the wvarious reactor concepts of interest,
synthesize that information in accordance with specific technical areas,
develop an understanding of how design features influence the overall
cost of generating power, and consider how changes in the design might
accomplish improved economic performance and acceptance by regulators
and the public. In addition to technical evaluation, assessments were
also made of the various nontechnical factors that influence commercial
use: for example, regulatory requirements, industry perspectives on
future technologies, public acceptance, electric power growth needs, and
economic conditions.

1.2 REPORT ORGANIZATION

This report is organized into four volumes. Volume I is the Execu-
tive Summary.10 Volume II, Reactor Concepts, Descriptions, and Assess-
ments (this volume) primarily describes and evaluates the various
concepts in accordance with a chosen methodology. The advantages and
disadvantages as well as needs for further R&D of each concept are
described. Volume III, Nuclear Discipline Topics,!! deals with generic
disciplinary issues relevant to nuclear viability and provides a more
detailed discussion of issues. Construction, economics, regulation,
safety and economic risk, nuclear waste transportation and disposal, and
market acceptance are considered both individually and as inter-
related. Acceptance of nuclear energy by utilities, regulators and the
public and institutional needs to revitalize the nuclear option are dis-
cussed. Volume IV is a comprehensive bibliography.12

1.3 REFERENCES FOR CHAPTER 1

l. Nuclear Power in America's Future, Atomic Industrial Forum, Inc.,
New York, New York, June 1984.

 

2. The Future of Electric Power in America: Fconomic Supply for Eco-
nomic Growth, DOE/PE-0045, U.S. Department of Fnergy, Washington,
D.C., June ]1983.
3.

10.

11.

12.

1-3

J. Skeer and D. Meyer, Two Views of the Nation's Electric Future:
A Comparison of Recent Studies by the U.S. Department of Energy and
the Congressional Research Service, DOE/PE-0047, U.S. Department of

 

Energy, Washington, D.C., February 1984.

"35th Annual FElectric Utility Industry Forecast, FEconomic Recovery
Drives kWh Sales Up," Electrical World, 49-56 (September 1984).

 

G. Samuels, The Outlook for Electricity Supply and Demand, ORNL/TM-
9469, Oak Ridge National Laboratory, Oak Ridge, Tennessee, April
1985.

M. Crawford, "The Electricity Industry's Dilemma,'
248-250 (July 19, 1985).

Science 229,

Nuclear Power in an Age of Uncertainty, Office of Technology

 

Assessment, 1984.

A. M. Weinberg et al., The Second Nuclear Era, ORAU/IEA-84-6(M),
Institute for Energy Analysis, Oak Ridge Associated Universities,
Oak Ridge, Tennessee, March 1984.

 

R. K. Lester et al., Nuclear Power Plant Design Innovation for the
1990s: A Preliminary Assessment, MIT-NE-258, Massachusetts Insti-

 

 

tute of Technology, Cambridge, Massachusetts, September 1983.

D. B. Trauger (ed.) et al., Nuclear Power Options Viability Study

 

Volume I, Executive Summary, ORNL/TM-9780/1, Oak Ridge National

 

Laboratory, Oak Ridge, Tennessee, September 1986.

D. B. Trauger (ed.) et al., Nuclear Power Options Viability Study
Volume III, Nuclear Discipline Topics, ORNL/TM-9780/3, Oak Ridge

 

 

National Laboratory, Oak Ridge, Tennessee, September 1986.

D. B. Trauger (ed.) et al., Nuclear Power Options Viability Study

 

Volume IV, Bibliography, ORNL/TM-9780/4, 0Oak Ridge National

 

Laboratory, Oak Ridge, Tennessee, September 1986.
2. GROUND RULES AND CRITERIA

2.1 GROUND RULES AND THEIR SIGNIFICANCE

To facilitate useful study, NPOVS concentrated on a carefully se-
lected, limited group of concepts by developing and applying ground
rules. The following three ground rules were selected:

l. The nuclear plant design option should be developed sufficiently
that an order could be placed in the 20002010 time period.

2. The design option should be economically competitive with environ-
mentally acceptable coal-fired plants.

3. The design option should possess a high degree of passive safety to
protect the public health and property and the owner's investment.

Ground Rule 1 determines the time period of interest. It is as-
sumed that, if orders of additional nuclear power are placed before the
year 2000, they will be filled by current or slightly modified designs,
primarily of LWRs. By the turn of the millenium, the anticipated demand
for power may permit consideration of advanced reactor concepts and
their associated advantages. For the present the concept must be sup-
ported by an active and capable industrial proponent with a current pro-
gram. It is considered very difficult, perhaps even impossible, for a
proponent to obtain funds, complete a design, conduct R&D, build a
demonstration plant or its equivalent, and demonstrate satisfactory
operation by 2010, unless design work is already underway. The concept
must have no major feasibility problems or major questions that must be
resolved by long-term, high-risk R&D prior to commercial acceptance.

The validity of this ground rule with respect to the need for new
electricity generating capacity in the time period 2000 through 2010 was
considered early in the study. Several referencesl™ were reviewed and
opinions were solicited from the government offices, industries, util-
ities, and universities that were consulted in the conduct of the
study. Although projections ranged widely, we concluded that need for
increased baseload capacity was sufficiently probable to justify the
chosen time period. More recently, a comprehensive study of the outlook
for electricity supply and demand was made at ORNL which further con-
firms the choice.® The summary of that report by G. Samuels is included
as Appendix B.

Ground Rule 2 states the obvious: For a concept to be viable, it
must be economical. The measure chosen is the most likely and perhaps
the only major alternative, the coal-fired power plant. Since the cost
of coal and its transportation vary widely with location, this ground
rule is somewhat site dependent. The most favorable situations for coal
might eliminate some or all of the nuclear concepts. However, other
problems such as mining, acid rain, and carbon dioxide buildup could
become dominant by the time period considered in this study.

2-1
2-2

Ground Rule 3: Although 1licensable plants are considered ade-
quately safe by NRC and the nuclear industry, passive safety provides
additional protection that is independent from engineered devices and
from human intervention or management. The added protection of the
owner's investment through passively safe designs may enhance the
acceptability of advanced concepts as viable power options. Passive
safety should help overcome intervenor's objections, public apprehen—
sion, and utility hesitation. These features also may simplify plant
operation for the owner-operator. Thus, passive safety, enhanced beyond
that of the present safety philosophy of primarily diverse and redundant
engineered (active) systems, may provide an ingredient to help
revitalize the nuclear industry.

2.2 CRITERIA

An early effort of the study was to develop criteria that reactor
designs would have to meet to become viable in the future. 1In the reac-
tor assessments, the seven criteria that were developed were used as a
guide to assess nuclear concepts. In most cases, lack of adequate and
detailed data necessitated the use of engineering judgment to determine
compliance with criteria. Often the judgment had to be supplemented by
the formulation of further R&D needs to facilitate more reliable conclu-
sions. These needs are identified for each of the concepts.

The criteria are augmented by a list of characteristiecs that pro-
vide further guidance for properties judged to be of importance to
nuclear power viability. The characteristics chosen are not readily
quantifiable but include features that complement and amplify the cri-
teria.

2.2.1 Listing of Criteria, Essential and Related Desirable
Characteristics, and Discussion of Their Applications

The criteria are listed below. Comments follow each criterion.

l. The calculated risk to the public due to accidents is less than or
equal to the calculated risk associated with the best modern LWRs.

This is a fundamental public safety criterion. To implement it
strictly, a probabilistic risk assessment (PRA) would be necessary
for each new concept and for the "best modern LWRs." This is a
desired situation; however, where PRAs are unavailable or inade-
quate, other approaches based on judgment must be used. Compliance
with this criterion is essentially a prerequisite for 1licensing.
The criterion indicates the future importance of PRA.
4,

2-3

The probability of events leading to loss of investment is less than
or equal to 10 per year.

Protection of the plant and its components and hence the in-
vestment of the utility is the focus of this criterion. PRA methods
must be applied again as for Criterion l, to which this criterion is
closely related. The emphasis is on the integrity of the entire
plant, starting with the reactor core. The stated probability of
10 per year for 1loss—of-investment events is numerically con-
sistent with insurance practices. Public reaction to nuclear acci-
dents may make a lower probability desirable, recognizing the con-
sumer pature of the utility industry sales. Although a probability
of 10 © has been cited as appropriate by other authors,l we believe
that 10 is a conservative figure for 1loss of plant investment
alone. A unique situation might arise if consideration were given
to the possibility that a class of reactors could be shut down in
response to a single faillure of one of the reactors in that class.
For the concept assessments, it was assumed that each plant stands
alone and is independent of all others. Thus, failure probabilities
are not nmultiplied by the number of units in a class.

The economic performance of the nuclear plant is at least equivalent
to that for coal-fired plants. (Financial goals for the utility are
met, and busbar costs are acceptable to public utility commissions.)

This criterion states the basic economic requirements for
acceptance of a power plant by a utility and a public utility com-
mission. It assumes that the available alternatives are primarily
coal-fired or nuclear power plants. The evaluation is based on data
provided by the proponent, checked for consistency, transformed to a
common basis, and compared with data for coal-fired power plants.
Purposely this 1is not a rigid criterion. It is recognized that
coal-fired plant costs vary with location, type and quality of coal,
and state and local regulations. Thus the competitive position of a
nuclear plant would depend on location as well as its own costs and
performance, but analysis at this level of detail is beyond the
scope of this report. This criterion also suggests the need for
better understanding of cost estimating, particularly cost savings
associated with replicas, shop fabrication, economy of scale, and
break-in of new technologies.

The design of each plant is complete enough for analysis to show
that the probability of significant cost/schedule overruns is
acceptably low.

This criterion also addresses the economic risk to the capital
investment as affected by unanticipated requirements and schedule
delays during the construction period up to the start of revenue-
producing operation. Sufficiently complete and detailed designs,
schedules, and specifications must exist to permit orderly planning
and to prevent or minimize unanticipated events that lead to cost or
S

6.

2-4

schedule overruns, An appropriate review would include the com-
plexity of design, requirements for accuracy and tight tolerances,
compactness of arrangements, room for expansion, and strictness of
sequencing requirements. Review by NPOVS is limited since designs
of the concepts considered are not complete enough for thorough
analysis.

Official approval of the plant design must be given by the NRC to
assure the investor and the public of a high probability that the
plant will be licensed on a timely basis, 1if constructed in
accordance with the approved design.

This criterion addresses concern for delays and associated risk
for fully designed or replica plants and is closely related to Cri-
terion 4. Criterion 6 also addresses the concerns of the adequacy
and sufficiency of the first plant. Although current regulations
provide a mechanism for the preapproval of standardized plant
designs (10 CFR 50, Appendices M, N, and 0; and 10 CFR 170.21) and
early site-suitability reviews (10 CFR 50, Appendix Q), there is
little experience in applying this mechanism. Also, past experience
shows that the existence of a completed and licensed plant does not
guarantee that a replica will encounter no obstacles in obtaining a
license., This criterion's prime concern is with the licensing pro-
cess, including potential further changes in requirements and regu-
lations. Experience with licensing is extensive and should be suf-
ficient to permit the induction of one-step licensing at the comple-
tion of design. Verification of quality control duriang construc-—
tion, of course, would be required. This criterion 1is also
addressed in the chapter on regulation in Volume ITI.

For a new concept to become attractive in the marketplace, demon-
stration of its readiness to be designed, built, licensed, and begin
operation on time and at projected cost is necessary. (The demon-
stration may be accomplished by a reactor dedicated to that purpose
or by a first-of-a-kind plant or the equivalent in other exper-—
ience. For the latter, it is necessary that all important features
have been demonstrated and that the sum of such features does not
constitute a major departure from existing licensing experience.)

For a concept to be seriously considered as a viable option in
the power iadustry, convincing evidence must be provided relative to
major economic and performance claims. A demonstration plant offers
an effective way to acquire this competence, Presumably the demon-
stration plant could be used for extensive validation of computer
codes related to safety and operability. It might directly demon—
strate selected safety features to the regulators, the industry, and
the public, The construction and start-up of the demonstration
plant would provide a base of experience from which future standard-
ized plants could be designed. It is possible that following the
test phase, the demonstration plant could be operated for an ex-
tended 1life as a first-of-a-kind unit. Where extensive related
2-5

prior experience is available, the probability for demonstration in
a first-of—-a-kind plant may be high.

7. The design should include only those nuclear technologies for which
the prospective owner/operator has demonstrated competence or can
acquire competent managers and operators.

For the operation of a new or substantially different concept
to be satisfactory, utility plant managers and operators must have
acquired an adequate background and experience with the technology
and equipment. This criterion relates closely to Criterion 6 since
the demonstration plant can provide an exceptionally good training
facility. Simulator training has proven effective for current power
plants, and simulators would be necessary tools for new concepts.
Where the concept derives from a prior system such as the small BWR,
this criterion should be relatively easy to meet.

The seven criteria have been selected as a primary basis for evalu-
ating new or existing reactor concepts. These criteria obviously are
not independent since criteria 1 and 2 deal with the probabilities for
successful operation or failure, criteria 3 to 6 are primarily economic,
and criterion 7 relates to operation. However, we deem each to have
sufficient stand—alone merits to justify separate consideration. The
four essential characteristics listed below in large measure amplify the
criteria. The desirable characteristics which follow are more per—
ipheral and, in some instances, are not applicable to all concepts; how-
ever, they provide a useful checklist for evaluation purposes.

The essential characteristics for future nuclear plants are as fol-
lows:

® Acceptable front-end costs and risks
— Construction economics

® Low and controllable capital costs (utilizing, for example,
shop fabrication, a minimum of nuclear grade components, and
standardization)

® Designed for long lifetime

— Investment economics including risk

® Low costs associated with accidents

® Low costs associated with construction delays

® Low costs associated with delayed or unanticipated actions by
regulatory bodies

® Low costs associated with delayed or unanticipated actions for
environmental protection

® Unit sizes to match load growth

® Uncertainties in technology and experience not 1likely to
negate investment economics
2-6

Minimum cost for reliable and safe operation

High availability

Minimum requirements for operating and security staffs
Designed for ease of access to facilitate maintenance
Simple and effective modern control system

Low fuel cycle cost

Adequate seismic design

Practical ability to construct

Availability of financing

Availability of qualified vendors

Availability of needed technology

Adequately developed licensing regulations applicable to the con-
cept

Ease of construction enhanced by design

Public acceptance

Operational safety of power plants

Safe transportation and disposal of nuclear waste and radiocactive
effluent _

Low effect on rates from construction and operation

Adequate management controls on construction and operation
Utility and regulatory credibility

Related desirable characteristics are as follows:

1.
2,
3.
b
5.
6.
7.
8,
9.
10.
11.
12.
13.
14,
15.

practical research, development, and demonstration requirements,
ease of siting,

load-following capability,

resistance to sabotage,

ease of waste handling and disposal,

good fuel utilization,

ease of fuel recycle,

technology applicable to breeder reactors,

high thermal efficiency,

low radiation exposure to workers,

high versatility relative to applications,

resistance to nuclear fuel diversion and proliferation,
on—line refueling,

ease of decommissioning, and

low visual profile.

Several of these characteristics are not readily determined quanti-

tatively and therefore are applied primarily by judgment. They indicate
areas and issues of interest and importance. As a rule, an individual
characteristic should not determine the fate or viability of a concept,
2~7

2.3 REFERENCES FOR CHAPTER 2

1.

2,

3.

4

5.

Re K. lLester et al., Nuclear Power Plant Innovation for the 1990s:
A Preliminary Assessment, MIT-NE-258, Massachusetts Institute of

 

Technology, Cambridge, Massachusetts, September 1983.

U.S. Department of Energy, Office of Policy, Planning and Analysis,
The Future of Electric Power in America: Economic Supply for
Economic Growth, DOE/PE-0045, June 1983.

 

Congressional Research Service, A Perspective on Electric Utility
Capacity Planning, Committee Print 98-M, U.S. Goveranment Printing

 

Office, 1983.

“35th Annual Electric Utility Industry Forecast,” Electrical World
(September 1984), pp. 49-56.

G. Samuels, The Outlook for Electricity Supply and Demand, ORNL/TM-
9469, Oak Ridge National Laboratory, Oak Ridge, Tennessee, April
1985.
3. CONCEPT EVALUATIONS

The NPOVS considered many councepts and selected for assessment
those that were judged to fit within the scope of the study. The
selected concepts are listed below. Exclusion does not necessarily
indicate that the concept can not be utilized effectively within or
before the time frame. The basis for exclusion has been either by the
Ground Rules or because as for the ABWR and APWR, the system is
sufficiently far along in commercial development that review here was
deemed unnecessary. Concepts considered but not chosen for further
evaluation are listed and discussed in Appendix C.

3.1 CONCEPT SELECTION AND CLASSIFICATION

All the concepts selected are considered advanced, are presently in
development and have various degrees of innovation as compared to
current concepts. For convenience, the selected concepts were clas-
sified in the traditional way by their coolant and respective generic
name.

The concepts selected are:
1. Light Water Reactors (LWR)

a. PIUS — Process Inherent Ultimate Safety — promoted by
ASEA~-ATOM

b. Small BWR — Boiling Water Reactor — promoted by General
Electric

2. Liquid Metal Reactors (LMR)

c. PRISM — Power Reactor Intrinsically Safe Module — the
General Electric advanced concept supported by
DOE

d. SAFR — Sodium Advanced Fast Reactor — the Rockwell

International advanced concept supported by DOE

e. LSPB — Large—-Scale Prototype Breeder — the EPRI-CoMO
concept supported by DOE and EPRI

3. High Temperature Reactors (HTR)
Side-by-Side Modular — The core and steam generator in
separate steel vessels in a side-by-side configuration. The

concept 1is supported by DOE and promoted by GCRA and in-
dustrial firms.

3-1
3-2

All the concepts selected are judged to fit within the scope of the
study. They are all potentially available in the time period 2000-2010.
All are estimated by their promoters to be economically competitive with
coal-fueled power plants, and attention has been given to designing them
to include features of passive safety.

3.2 CONCEPT EVALUATION METHODS AND LIMITATIONS

The assessment of the concepts selected was performed using the
criteria as a guide. First, available information has been gathered and
is summarized in this report. Claims, advantages, and disadvantages
were then formulated for each concept. In many cases, the available
information is not sufficient to reach a final conclusion. Need for
further information is then formulated in the form of R&D requirements
including design.

An opportunity was given to the concept proponents to comment on
the material related to their concept. These comments were considered
carefully, but not all were accepted or included in the text. The pro-
ponents’ assistance has been highly valuable and is appreciated.

The descriptions and evaluations are primarily based on information
available in September 1985. Since all of the concepts have continued
in development, design changes may affect the information presented
here. Although the observations of the report are considered valid at
the time of writing, the reader should use care in determining subse-
quent design changes and in assessing their impact before drawing ser-
ious conclusions.

3.3 LIGHT WATER REACTORS (LWRs)

The next two sections discuss the light water reactor (LWR) con-
cepts investigated by NPOVS. The Process Inherent Ultimate Safety
(PIUS) and Small (Advanced) Boiling Water Reactor (BWR) concepts are in
preliminary stages of development. Both concepts build on LWR experi-
ence.

3.3.1 PIUS

 

3- 3. 1.1 DescriEtion

The Process Inherent Ultimate Safety (PIUS) concept!™ is basically
a PWR-type NSSS with core reactivity controlled both by varying the
soluble boron conceuntration in the primary system coolant circuit and by
the natural feedback provided by the fuel and moderator negative temper-
ature coefficients of reactivity. The current “modular” plant design"
consists of three separate cores located within three separate primary
circuits which are surrounded by a common pool of heavily borated water
3-3

maintained at low temperature. The pool and the three reactor-steam-
generator modules are contained within a prestressed concrete pressure
vessel (PCPV) and utilize a common pressurizer. Steam is routed from
each steam generator to a conventional balance of plant consisting most
likely of a common steam header feeding one or possibly two turbine-
generator sets.

The reactor system is designed to provide natural thermal convec-
tion under reactor shutdown conditions by placing each reactor core at
the base of a hot water column about 30 meters high. A large opening is
provided between the primary system in each module and the pool water at
both the top and bottom of the core riser column. Honeycomb arrays of
tubes are provided in these open regions to inhibit convection currents
during normal operation of the reactor, with the hydraulic head of the
fluid in the surrounding pool being balance by the pump head in each
primary circuit. Upon loss of forced convection flow in one of the pri-
mary circuits, the difference in density between the hot and cold water
regions causes the borated pool water to flow into the core region of
the affected module.

Figure 3.1 illustrates the axial configuration of a single reactor-
steam—generator module showing the flow paths of the primary circuit,
The PIUS Mk II module, illustrated in Figure 3.1, utilizes a single
chimney~type riser pipe to carry coolant from the core up to a hot leg
recirculation pump from which primary circuit flow proceeds down through
a bayonet steam generator in a single pass to the core inlet. Density
locks are provided between the primary circuit and pool water both below
the core and around the pressurizer riser pipe from the primary
recirculation loop. Figure 3.1 shows the connection between the module
and the common pressurizer for the three modular wunits. Figure 3.2
provides two plan views of the intermnals of the PCPV showing the fuel
storage locations and the mobile refueling turntable below the reactor-—
steam—generator modules on the right and the upper piping connections on
the left., Figure 3.3 illustrates the PCPV with a central cavity and a
drawer-1like closure slab and pressure seal.

The system is designed for a constant water flow rate through each
primary circuit and a constant reactor outlet temperature over the
normal range of power operation. In the event of pump stoppage in any
or all modules, the forced convection coolant flow through the primary
circuit pump and steam generator comes to a stop, but thermal convection
through the core and into the pool will initiate because of the differ-
ential head available for natural thermal convection. This causes the
cool borated pool water to enter the core, quenching the nuclear reac-—
tion and cooling the fuel elements. Shutdown initiated for a single
module should not affect operation of the remaining modules. The volume
capacity of the PCPV should be sufficient to provide cooling water for
about a week to accommodate decay heating from all three reactors
without addition of water from an external source.

The concentration of boron within each core region is controlled by
injecting deborated <coolant obtained from one of two separate
3-4

ORNL—-DWG 85-13160R

 

  

 

 

A
PRESSURIZER STEAM VOLUME

 

 

  

REACTOR VESSEL CLOSURE

 

 

 

 

 

 

 

 

 

PRESTRESSED CONCRETE UPPER HOT/COLD INTERFACE

REACTOR VESSEL (PCRV)

 

RETURN FLOW PIPE

 

 

 

 

STEAM PIPE :

7 .
FEEDWATER - z z
PIPE

~,

 

SCRAM VALVE

 

 

 

 

 

 

 

 

 

 

 

 

STEAM GENERATOR
TUBE BUNDLE

POOL

 

 

 

 

 

 

DOWNCOMER

 

 

 

RECIRCULATION PUMP MOTOR
RISER PIPE

 

 

 

 

 

 

 

 

 

 

 

PLATFORM (REVOLVING)

 

 

 

CORE

CORE SUPPORT
PLATE

 

 

 

 

 

 

 

 

 

 

 

 

N\

 

 

 

 

LOWER HOT/COLD ‘ -
INTERFACE (i

~7,

 

 

 

s
A
ey

Fig. 3.l. Vertical section through one module of a MK IT modular
PIUS plant.
ORNL-OWG 85-13158

RECIRCULATION

  
 
 
  
  
   

  
  
 
    

STEAM GENERATOR

(UPPER END) clikiee ROTATABLE

PLATFORM
SEISMIC
RESTRAINT

INTERIM FUEL
STORAGE RACK

 

’/"_REACTOR MODULE
IN OPERATING

POSITION
DENSITY LOCK

FUNNEL

 
    

e -
o .

e w . .
’ cala @ Q@

 

£:<
SPENT FUEL RACKS

   
 

Fig. 3.2. Horizontal sections through a three-module MK IT PIUS plant.
3-6

ORNL-DWG 85-13159

 

 
   

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

¥ [ SR LTNLTY

 

 

 

 

 

 

 

 

 

 

 

 
  

 

 

 

 

 

 

 

 

 

 

 

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3-7

distillation units. Both distillation units may be used after a scram
to reestablish plant readiness. The size of the collection tanks and of
the borated and clean water tanks 1is such that two consecutive scrams
can be accommodated with only one evaporator operating. Each reactor
will operate without control rods, and the boron concentration control
will provide the necessary reactivity control above that provided by in-
herent reactivity feedback and burnable poisons,

The PIUS reactor concept is proposed as a very safe reactor uti-
lizing LWR technology. Further, it is proposed as a system which has a
competitive capital cost because of its safety and control character-
istics. The practicality of the concept rests on the stability and con-
trol of the fluid interfaces between the pool water and each hot primary
circuit, including stability under transient conditions. Other impor-
tant factors are the practicality of the PCPV design, of refueling by
use of a rotating turntable within a deep pool, of insulation require-
ments, of replacement of equipment, the non-safety grade balance of
plant and the practicality of operating three reactors without any
control rods and without interference due to thermal and pressure
transients. All of the above areas are being evaluated.

A fundamental design goal of PIUS is the preservation of fuel
integrity under all credible conditions. The approach is claimed to
achieve complete protection against core melting or overheating due to
the following:

- any credible equipment failure,

- mnatural events such as earthquakes and floods,
- reasonably credible operator mistakes,

- combinations of the above
as well as

- 1inside sabotage by plant personnel,
- terrorist attacks in collaboration with insiders,

- military attack (e.g., by aircraft with 'off-the-shelf' non-
nuclear weapons),

- abandonment of the plant by operating personnel.

The designers state that every attempt is made to achieve these
goals via passive means (i.e., without reliance on safety equipment
which could fail to function). This passive protection should last for
a minimum of one week following any initiating event. Thus PIUS should
meet well the Third Ground Rule. The designer's position is that all of
these goals can be attained via fulfillment of two basic safety func-
tions:

1. the design precludes any credible circumstance which can result
in uncovering of the core, and
3.3.1.2

3-8

the design prevents core heat generation rates which would ex-
ceed the convective heat removal capability of the submerging
water under all conceivable conditions.

Claims, Advantages, and Disadvantages Evaluated Against

 

Criteria, Essential and Desirable Characteristics

 

The proponents claims and potential concept advantages are dis-
cussed briefly as follows first in the order of the preliminary criteria
and then with regard to both essential and desirable characteristics:

1.

3.

Public Risk: PIUS appears to be resistant to fuel damage
under all anticipated traansients and for long periods (~7-10
days) without human intervention or active engineered safety
features after worst case accidents. ASEA-ATOM claims that
normal operating releases are expected to be less than for
current LWRs. The 1lower power density and expected good
chemistry behavior should limit the occurrence of minor fuel
leaks. ASEA—-ATOM argues that pressure cycles in the large
pool where depleted fuel is stored will not exceed clad
strength limits so that pool contamination will be minimal.
Steam generator tube leaks will be contained by steam—line
isolation valves and a secondary pressure boundary rated at
primary system pressure (9 MPa or 1300 psia) up to the
isolation valve location just outside the wall of the PCPV.
PIUS is also designed to be highly resistant to external
threats such as aircraft crashes, terrorism, and sabotage.

Investment Risk: Since core damage is precluded during normal
transients and delayed for at 1least a week without human
intervention during worst case accidents, the probability of
loss of investment arising from core melt appears to be much
less than 10™%/plant year (no PRA yet to support this con-
clusion).

Economic Competitiveness: ASEA-ATOM claims economic competi-
tiveness with coal-fired plants under Swedish market condi-
tions and assuming a aon-safety grade ©balance of plant
(BOP). Their unpublished studies show that a PIUS 500 MW(e)
nonmodular plant has twice the capital cost of a Swedish 600
MW(e) coal-fired plant but an energy cost that is less than
the coal-fired plant by between 15% (for coal at $50 per
metric ton) and 25% (for coal at $60 per metric ton), where
the cost of coal contributes approximately 60-70Z of the
levelized energy cost. Similarly, recent ASEA-ATOM estimates
reported for the 600 MW(e) modular plant show the same rela-—
tion to coal-fired capital costs and between a 20% and 307 ad-
vantage in energy costs for the coal prices noted above.
ASEA-ATOM has not provided supporting documentation of the
analysis nor 1is there sufficient information available cur-
rently to perform an independent analysis. However, if the
4.

3-9

PIUS can be built at twice the capital cost of a coal-fired
plant, it would be competitive in the United States in regions
where Western low=-sulfur coal is selling at $35 per metric ton
and where Eastern high-sulfur c¢oal is selling at $45 per
metric ton assuming that the price of coal in the future
escalates at 1.5% or more above inflation.

Probability of Cost/Schedule Overruns: ASEA-ATOM recognizes
the need to develop a complete design before initiating con-
struction and is working toward that goal.

Licensability: ASEA-ATOM has a draft 1licensing plan, is
actively engaged in dialogue with NRC-NRR and has plans to
proceed with securing NRC final design approval (FDA) by
January 1992.

Demonstration of Readiness: ASEA-ATOM believes that a non-
nuclear demonstration plant (including a full-scale steam
generator module) built in a tank to demonstrate the
fundamental thermal-hydraulic safety principles of PIUS can
demonstrate the resistance of the plant to wunwarranted
shutdowns from minor transients. They further believe that
such a demonstration will simplify licensing and quell the
arguments of detractors with regard to the potential of high
unavailabilities resulting from minor upsets. ASEA-ATOM also
believes that a non-nuclear demonstration presents the
possibility of going directly to a financially selfsupporting
power plant without burdening the development program with the
large expeunditure and time delay of a nuclear demonstration.
ASEA~ATOM believes that the full-scale non-nuclear test would
resolve most constructibility problems except for the PCPV;
however, they consider the constructability of the PCPV to be
based on an established technology and to be easier to con-
struct than their curreat BWR pressure suppression containment
buildings. Such a demonstration would not fully address all
issues of constructability and maintenance, particularly long-
term requirements for steam generator cleaning.

Owner Competence: Any previous operator of an LWR should be
able to build and operate PIUS assuming that the plant
operates as projected by ASEA-ATOM., The operational complica-
tions are possibly the use of three reactor-steam generator
modules to feed a single steam header, steam generator tube
cleaning, pressure control of the primary system and refueling
three cores in a deep pool. ASEA-ATOM claims that these fea-
tures actually simplify operation and that their analyses sup-
port this claim. ASEA-ATOM points to the use of the large
(~200 m3) pressurizer as a buffer during transient events.
Actual operating experience 1is needed to corroborate these
claims. Other technologies appear to be extensions or modifi-
cations to existing IWR technology.
3-10

Essential Characteristics: PIUS has great promise according
to its proponents of achieving safe, reliable and cost com-
petitive operation; however, as noted under the evaluation
against criteria, most of these claims require further anal-
ytical or experiential substantiation, particularly in the
area of construction costs and availability. The sizing of
PIUS at about 600 MW(e) appears to meet the need for smaller
increments in base load capacity additions consistent with
current load growth projections, and modularization of the
reactor-steam generator configuration is an attempt to
simplify plant construction. Resistance to accidents and
external events such as sabotage is inherent in the design as
conceptualized by the proponents. Such resistance implies low
costs associated with accidents, possibly smaller security
staffs and the reliance on passive safety features under worst
case scenarios to delay required mitigative actions and
perhaps to avoid having to provide for area evacuation.

Desirable Characteristics: Aside from the lack of planning
for fuel recycle, the low thermal efficiency, lack of on-line
refueling, and the questionable versatility relative to appli-
cation (except for district heating) because of the low grade
of steam compared to other higher temperature concepts, PIUS
possesses most of the desirable characteristics listed in
Chapter 2. Some areas of potential weakness such as waste
handling and disposal, decommissioning and diversion and
proliferation are generic problems shared in common with vir-
tually all reactor concepts because the proposed solutions to
these type of concerns are often subject to a variety of
interpretation by national and international regulatory,
governmental and public interest bodies. Replacement of the
nuclear steam supply equipment for extended life may be easier
for PIUS than for other LWR concepts assuming that the vessel
remains qualified. Because of passive safety characteristics,
PIUS does appear to be potentially much more flexible to
siting requirements and more much resistant to sabotage and
diversion than current generation LWRs.

The potential disadvantages are discussed as follows:

1.

Public Risk: The potential for refueling accidents must be
addressed more fully with attention to the control and moni-
toring of heavy loads above the spent fuel storage locations
within the PCPV. ASEA-ATOM claims that major maintenance such
as pump replacement could be performed with the freshly
exposed fuel in the operating positions under the steam gener-
ators. The possibility of damage to fuel stored from previous
operating cycles, claimed to pose little radiological hazard,
may need extensive study.
3~11

Investment Risk:

de

b.

PIUS availability has not been firmly established by
analysis and testing to date.

Deep pool maintenance may be a problem with the onus of
long shutdowns. ASEA-ATOM cites their replacement of BWR
internal recirculation pumps as an adequate experience
base for deep pool maintenance on difficult configura-
tions.

Economic Competitiveness:

2

b.

d.

€.

PIUS competitiveness with coal and other nuclear options
has not been established independently. PIUS capital
costs have not been confirmed for U.S. siting.

On-line refueling is not possible for PIUS so refueling
shutdowns must be scheduled as in current IWRs. The
three cores must be shut down for one to be serviced.

PIUS load following may have been enhanced by a three
reactor system, but operation of one or more modules at
reduced power may be difficult depending upon the diffi-
culties encountered in controlling steam generator feed-
water injection rate to match core power levels. The
effects of fuel cycle cost penalties have not been ade-
quately addressed for the case in which the three cores
become out of phase in burnup. ASEA-ATOM states that
less fuel will be added in refueling the affected module;
this solution to get back in phase incurs a financial
penalty which is acknowledged but not fully addressed.
Another alternative 1is to interchange fuels between
modules.

PIUS is not amenable to efficient fuel recycle because of
the emphasis on onsite 30-year storage of spent fuel.

PIUS has a relatively low thermal efficiency.

Probability of Cost/Schedule Overruns: No specific disadvan-

tage

identified provided component and system testing have

been completed successfully prior to design completion.

Licensability:

de

As with all advanced reactors, PIUS has not yet been
forced to address post-TMI licensing proceedings. The
possibility exists for the need to address the probabil-
istic risk of the consequences of beyond design basis
accidents coupled to the cost-benefit of adding equipment
to avert the risk associated from very low probability
accidents. The regulatory authorities may require the
3-12

use of a separate containment, a dedicated 30-day water
supply, control rods, a safety—-grade control room,
dedicated emergency electrical power and/or other safety
grade systems on the balance of plant. 1If so, the effect
will be increased plant cost to meet such requirements.
There appears to be a potential need for some level of
secondary containment during shutdown operations with the
vessel seal broken.

b. Requirements for area evacuation are also not firmly
established for PIUS. ASEA-ATOM argues convincingly that
significant releases could only occur after an exteaded
time delay (7-10 days) following an accident. Such a
delay in taking action to add water, assuming adequate
prior provision had been made, would imply a chaotic
social situation external to the plant. However, the
PIUS pool could become severely contaminated from unan-
ticipated fuel failures, for example, due to manufactur-
ing problems. The result would be either a long shutdown
to allow decay of radioactive nuclides with relatively
slow water cleanup rates or an accelerated cleanup rate
in which out-of-reactor incidents would be possible.
ASEA-ATOM argues that eliminating the need for immediate
evacuation is being addressed and that their water clean-
up systems should be comparable in speed and safety to
current IWRs and this appears to be achievable.

Demonstration of Readiness: Steam generator cleaning require-
ments may not be confirmed until after the first plant has
been in operation for some time. ASEA-ATOM argues that clean-
ing requirements can be studied in relatively small scale
(non-nuclear) tests, but such idealized testing although im-
portant may be inadequate to represent actual plant opera-
tion. Improvements in design for cleaning have been reported
by ASEA~ATOM, but details are not yet available. Fuel
handling, in-pool maintenance, boration control, wet insula-
tion, submerged pumps, core carrier manipulation, and the PCPV
closure involve new designs and technology. Although these
features can be tested in mockup facilities, they will require
actual demonstration with reactor operation. However, since
the concept draws heavily on PWR experience, the demonstration
reactor may be determined as commercially viable following a
successful test period.

Owner Competence: No specific disadvantage identified.

Essential Characteristics: Most of the significant potential
disadvantages associated with the essential characteristics
have been addressed above under the criteria for Investment
Risk, Economic Competitiveness and Licensability. As men-
tioned under the discussion of advantages, estimates of cost
and availability need to be substantiated.
3.3.1.3

3-13

Desirable Characteristics: Most of those not met by PIUS were
listed above for comparison purposes in the evaluation of the
advantages. Load following is one area which the PIUS propo-
nents claim has been solved by modularization but the control
system, as noted under the economic competitiveness, may be
more complicated and harder to operate effectively than cur-
rently envisioned by the proponents. However, experience with
control in experimental facilities is reported to be favor-
able. With respect to radiation exposure to workers, corro-
sion product activation and transport 1is an unresolved
consideration since the behavior may be different from that of
the standard PWR.

R&D Needs and Open Questions Evaluated

 

Development, testing and demonstration of an optimum geometry
for the hot/cold water interface mechanism to ensure core
shutdown/quench when required and to preclude unwarranted
shutdown 1in response to minor transients and upsets are
needed. ASEA-ATOM has conducted preliminary tests which pro-
duced favorable results, but more comprehensive testing and
analysis are required and are planned.

ASEA-ATOM is attempting to demonstrate by analysis and testing
that plant availability will be sufficiently high to avert the
possibility that the design is too complicated to operate
economically. This important evaluation would justify
independent study.

Safety relief valves, steam suppression and filtering systems
may require extensive testing since they are essential compo-
nents of the decay heat removal system.

Components within the PCPV, the unusually long teandons, the
vessel liner, and its closures must be carefully studied and
designed to ensure that in-service inspection is practical
where required.

Technical, economic and licensing evaluation and assessment is
necessary for the PCPV including the sliding upper cover and
locking devices.

Submerged steam generator development is required including
single tube, multitube and fullscale development testing of
flow stability and transient response; disassembly, cleaning
and plugging procedures.

Wet thermal insulation for the primary coolant system and PCPV
requires further development and demonstration testing.

Development and testing of the refueling turntable component
handling tools, and other features of underwater maintenance,
3-14

are needed. Also, corrosion product transport and activation
are likely to differ from the standard PWR and may complicate
maintenance.

9. Economic evaluation must be performed comparing against coal
and conventional LWRs. This evaluation must include a capital
investment cost estimate for U.S. siting. Planned and real-
istically estimated forced outage rates must be evaluated
against refueling and maintenance requirements and the spectra
of potential plant unavailabilities. Extended testing of com-
ponents is recommended in borated coolant conditions.

10. Planned outage rates and realistically estimated forced outage
rates must be evaluated against refueling and maintenance
requirements and other potential plant wunavailabilities.
Extended testing of components is recommended under borated
coolant conditions.

11. Development and demonstration testing are needed for the
multi-service pressurizer and the integrated control system

for three reactor-steam generator modules feeding one or more
turbine—-generator sets.

12. Confirmatory tests and analysis will be required to assure

adequate negative reactivity feedback and the effective
operation of the soluble boron control system.

3.3.2 The Small Advanced BWR

 

3.3.2.1 Description

A small Boiling Water Reactor (BWR) design concept5 has been devel-
oped by the General Electric Company (GE). This concept attempts to
maximize the use of BWR design, technology, and operating experience.
Significant innovations are included to simplify and improve the per-
formance of safety functions. These, as well as other system simplifi-
cations and a reduced power rating, are claimed to reduce total costs
and speed construction. The major emphasis by GE has been on a
600 MW(e) concept which is judged to be adequately competitive with coal
to interest the U. S. market. Lower power ratings are possible. A
1000 MW(e) concept is feasible but would require a larger volume sup-
pression pool to maintain an equivalent time for operator response as in
the 600 MW(e) concept.

The small BWR concept (Fig. 3.4) uses an isolation condenser to
improve transient response. Control rods, which can be driven either
electrically or with accumulator pressure, and gravity-driven borated
water injection from an elevated low pressure pool are used to simplify
and provide diversity to the shutdown function. Core cooling and decay
heat removal are provided by depressurizing the reactor to the elevated
UPPER
STRUCTURE
' RELIEF

3-15

ORNL-DWG 85-9578

 

 

 

 

CONTAINMENT
STRUCTURE

 

 

ISOLATION
CONDENSER -

 

HORIZONTAL

VENTS

DRYWELL” =

 

 

 

 

 

 

 

 

 

   

 

 

 

REACTOR CORE ——

INTERNAL

   

 

BOTTOM

CONTAINMENT
OVERPRESSURE
RELIEF

DEPRESSURIZATION
VALVES

 

 

SAFETY
VALVES

 

 

 

UPPER
STRUCTURE

LOW PRESSURE
LONG-TERM
MAKEUP

 

 

 

SRR

LOW PRESSURE
| ELEVATED POOL

N

b

 

NS

 

    

GRAVITY

 

 

 

FEEDWATER LI

 

 

 

STEAM LINE

 

MOUNTED
DRIVES

Figo 3.40

 

STEAM INJECTOR

A small BWR concept.

DRIVEN

EMERGENCY CORE
COOLING AND
LIQUID POISON BACK-UP

NE
3-16

suppression pool. The drywell and pool gas spaces are inert. In addi-
tion, a steam injector is used to improve feedwater availability by pro-
viding a continuous minimum flow from a condensate storage tank even if
the main feedwater pumps are lost.

Steam is produced in the reactor vessel in a manner similar to that
of current BWRs. 1Internal recirculation pumps similar to those used in
ABWR are wused to circulate water through the core. (At lower power
levels, natural circulation becomes a better choice.) The steam—water
mixture exiting the core 1is directed to separators and dryers.
Bottom-mounted control rod drives, similar to those used on the ABWR,
are used to provide power shaping and emergency shutdown.

Reactor pressure is normally controlled with turbine throttle and
bypass valves. When the reactor vessel is isolated from the turbine
condenser, an isolation condenser controls pressure. This device was
selected because of its simplicity and because it provides high pressure
reactor water 1inventory control. Failure of the isolation condenser
function (to control reactor pressure) is not expected during the plant
life. The availability of the pressure control function is to be
achieved by selecting an appropriate redundancy in the isolation con-
denser units and by using the safety valves/steam injector as the di-
verse backup pressure/inventory control system. However, if such a
failure occurs, safety and depressurization valves provide a backup de-
pressurization to the suppression pool which is positioned above the re-
actor vessel. When the reactor pressure is sufficiently low, check
valves open in the suppression pool-to-vessel fill lines and water flows
by gravity into the reactor vessel to keep the core covered. The re-
sponses to a loss-of-coolant accident and a transient with failure to
scram are similar.

The suppression pool contains horated water to provide a diverse
backup to the control rods. Core cooling and decay heat removal is as-
sured with water returned to the reactor vessel and steam produced by
decay heat is vented to the suppression pool. In the 600 MW(e) concept,
there is a three-day supply of water available to accept decay heat. No
operator action is required during this time. For longer periods the
suppression pool must be refilled from an assured source with highly re-
liable equipment. Emergency diesel generators and core cooling pumps
are not required.

With the above-mentioned safety features, a severe accident is ex-
tremely unlikely. The ability to retain fission products in the sup-
pression pool is an important BWR feature which has been retained to
provide mitigation of severe accidents. Use of simple safety devices,
activated by stored energy and use of inherent processes such as natural
circulation and gravity-fed water delivery to the core, reduce costs
through modularization and system elimination. The licensing process
may be simplified.
3-17

The following objectives were established by GE for the Small BWR
concept to assure that it would be practical and that enhanced perfor-
mance, economy, and safety would be achieved.

1.

The major power—-producing elements of the concept are based on
proven technology or minor extensions of current technology.

The key safety functions are maintained at all times during
transient and accident conditions:

Ae

Ce

There is no need for short-term operator action. Short-
term is defined as three days. After this period, opera-
tor actions which are required should be judged easy to
accomplish. With such operator action, the safety func-
tions will be maintained for an indefinite period.

Safety devices or features are either inherent to the
concept or rely on the use of stored energy for motive
power.

Extensive testing 1is not required. This 1is defined as
taking less than three years to prove the concept by
testing its new features. The first commercial applica-
tion 1s to serve as a demonstration unit, and a unique
demonstration is not to be required.

The design is capable of modularization to allow factory fab-
rication and testing of most major components.

The design permits a plant construction period of four years.

Capital costs are minimized so electricity generation costs
are competitive with those of coal-fired plants of similar
power ratings.

These concept objectives were established by GE with the goal of achiev-
ing high confidence that the final design will produce the required
power in a manner which enhances safety in a readily demonstrable way so
that the licensing effort can be simplified.

3.3.2.2.

Claims, Advantages, and Disadvantages Evaluated Against

 

Criteria, Essential and Desirable Characteristics

 

The proponent's claims and councept advantages are discussed briefly
as follows first in the order of the criteria and then the character-

istics:

1.

Public Risk: The Small BWR appears to be resistant to fuel
damage under all anticipated transients and for long periods
(3 days) without human intervention or active engineered
safety features after worst case accidents.
3-18

Investment Risk: The small BWR appears to have a high degree
of core damage resistance. The probability for loss of in-
vestment due to core melt would appear to be less than 1074/
plant year (GE claims to have a preliminary PRA to support
less than 10~®/plant year). The probability of high avail-
ability through resistance to long shutdowns from upsets or
major maintenance appears to be favorable for the small BWR.
This confidence derives in part from the similarity to current
BWRs and their extensive base of operating experience. The
small BWR has no external recirculation piping to be subject
to stress corrosion cracking, but a boration transient will
require deborating the primary system to acceptable levels.
GE 1intends to ensure that this transient will be a very low
probability event (less than once in plant life).

Economic Competitiveness: Based on GE studies to date, the
Small BWR is estimated to be nearly competitive with coal at
600 MW(e); however, GE believes that there are as yet un-
explored options to improve economic competitiveness. How~
ever, the design 1is too preliminary for definitive evaluation
of construction costs. Fuel cycle costs should derive easily
from past BWR experience as should plant availability.

Probability of Cost/Schedule Overruns: No specific advantage
identified.

Licensability: The licensability of this concept is enhanced
by its similarity to current BWRs and by GE's plans to take no
exceptions to the General Design Criteria. For example, the
concept includes the use of containment and control rods.
This approach is expected to minimize potential licensing dif-
ficulties.

Demonstration of Readiness: GE believes that the passively
safe BWR represents only a small evolutionary extension of
existing technology needing only about two years for R&D
demonstration testing and another two years to complete the
design. R&D costs are estimated by GE at about $3M but this
figure may be low. Their total design development cost is
roughly between $100M and $300M. The first plant may be ac-
ceptable for commercial operation after serving as the demon-
stration. GE believes that because the concept relies heavily
on existing technology and already developed designs, there is
no need for long term, extensive testing. Therefore, a plant
of this type could be available earlier than the 2000-2010
time frame.

Owner Competence: The small BWR should be amenable to ready
ease of operation by experienced BWR owner/operators.

Essential Characteristics: As in the case of PIUS, the small
BWR has a good deal of promise to provide safe, reliable and
3-19

economic electrical power. Unlike PIUS, the small BWR can be
compared more directly to its evolutionary antecedents in the
large BWR. From the standpoint of characteristics affecting
constructibility, cost and operability, the comparison breaks
down because of institutional reasons, for example, the his-
tory of problems in the U.S. BWR experience compared to suc-
cesses in the Japanese BWR industry. GE recognizes this dif-
ference in historical perspectives but offers no specific ar-
guments that the sought after characteristics 1listed in
Chapter 2 will be realized for the small BWR in the U.S. 1In
principle, at this early stage of development, most of the
essential characteristics can be ascribed to the small BWR
concept because of its enhanced reliance on passive safety and
design simplification; however, the elevated pool may impose
strenuous seismic requirements and more complicated construc-—
tion which can increase cost.

Desirable Characteristics: Because of the evolutionary nature
of the design, the few and practical RD&D requirements are the
most salient characteristic.

potential disadvantages are discussed as follows:
Public Risk: No specific disadvantage identified.
Investment Risk: No specific disadvantage identified.
Economic Competitiveness:

ae The cost competitiveness of the small 600 MW(e) BWR is
not yet demonstrated because the GE analysis, which is
stated to be very preliminary, finds the lower rating of
the small BWR somewhat short of breakeven with equivalent
coal-fired units. When other potential cost savings are
anlyzed, GE expects the design to be competetive with
coal. GE agrees that more detailed calculations are in
order and that independent evaluations are desirable.

b. On-line refueling is not possible for the small BWR, and
so refueling shutdowns must be scheduled as in current
LWRs .

Probability of Cost/Schedule Overruns: No specific disadvan-
tage identified; however, the design is in a very preliminary
state. TIdentifiable important needs which require attention
are a suitable water level indicator and a reliable depres-
surization system.

Licensability:

ae As with all advanced reactors, the small BWR has not yet
been forced to address post-TMI 1licensing proceedings.
3-20

However, GE is aware of post-TMI requirements and plans
to reflect them in the final design. The possibility
exists for the need to address the probabilistic risks of
the consequences of beyond design basis accidents coupled
to the cost-benefit of adding equipment to avert the risk
associated with such very low probability accidents. GE
feels that retention of the suppression pool and its
ability to retain fission products after a severe acci-
dent will be of benefit in analysis of such accidents.
The regulatory authorities may impose the use of second-
ary plant man-rated shielding for power operation, a
dedicated 30-day water supply for the suppression pool,
enhanced safety-grade control room beyond that planned by
GE, and/or dedicated emergency ac electrical power. GE
intends to rely solely on emergency dc sources with the
gravity drain ECCS and argues that diesel-powered pumps
on a fire engine truck could provide water to the sup-
pression pool after 3 days. Although GE acknowledges
that there are some licensing risks with potential for
cost iancreases, GE also contends that this potential ap-
pears relatively small.

b. Need for area evacuation also is not firmly established
for the small BWR, but GE feels that they can make a good
case for avoiding area evacuation based on expected per-
formance at lower power density and with expected fission
product retention in the suppression pool.

Demonstration Plant: No specific disadvantage identified.
Owner Competence: No specific disadvantage identified.

Essential Characteristics: Although adequately safe by cur-
rent design standards and regulations, BWRs have been signif-
icant contributers to radioactive effluent releases. Better
fuel-clad performance and improved water chemistry have sub-
stantially decreased the releases. The use of barrier fuel
and lower power densities as proposed should further reduce
the radicactive effluents. Construction of the elevated pool
and the attendant concern about seismic response must be ad-
dressed fully as planned in the proposed R&D.

Desirable Characteristics: The small BWR offers no new fea-
tures with regard to on-line refueling, proliferation re-
sistance, decommissioning or waste handling and disposal. The
concerns and problems are comparable to current generation
LWRs and in many instances shared with other advanced or inno-
vative designs. Although the objective is to have a remotely
controlled secondary plant with no on-station operators near
potentially contaminatable steam lines, the small BWR shares
the same characteristics of the large BWRs in having radio-
activity in the steam lines which can lead to the potential
3.3.2.3

3.

3-21

for higher occupational exposures compared to other re-
actors. The improvements described in Section 8 above also
apply here.

R&D Needs and Open Questions Evaluated

 

A full height demonstration test is needed of the gravity-
drain Emergency Core Cooling System (ECCS) from the elevated
low pressure pool (thought to be an item of concern to NRC
licensing).

High flow steam injector demonstration testing is necessary;
the design is based on low flow tests performed over 10 years
ago. Also, the reliability and availability of the injector
control system must be demonstrated. A key component of the
injector control system is a sufficiently accurate vessel
water level indication. GE contends that injector control
does not require fine indication of vessel water level, which
may well be the case, but demonstration testing is felt to be
required to assure that the injector is a legitimate benefit
to plant safety aand enhanced availability. A potential buyer
may require assurance before being sold a product that could
cause delay, rework or even removal if it were to fail during
hot functional tests on the completed plant. GE notes however
that "in the worst case"”, the steam injector could be deleted
from the design, and a nearly equivalent level of protection
could be added by increasing the redundancy of the isolation
condenser, or by reverting to the use of the steam driven
Reactor Core Isolation Cooling (RCIC) system now in use at
operating BWRs.

Confirmation of the analytical load definition for horizontal
vent discharges to a covered suppression pool is needed. The
covered pool results in the need for load definitions up to a
differential pressure of about 0.35 MPa (40-50 psi) compared
to the currently available test data up to 0.1 MPa (15 psi).
However, such testing is already planned for the advanced BWR
in 1986, and the results of that program can be utilized for
the 600 MW(e) plant. GE considers this work to Dbe
confirmatory rather than R&D and so does not factor this into
their R&D cost estimates.

Depressurization valve demonstration is needed; the design is
based on a concept of a valve which is deenergized to open and
then latches open. Although the design is simple in concept,
it apparently has never (or not widely) been applied in prac-
tice. Most LWRs use electric or air operated valves which are
typically designed to fail shut; in the small BWR, reactor
pressure opens the valves against a magnetic force which is
lost when the dc¢ circuits are deenergized. GE considers air
operated valves to be the best backup alternative to the mag-
netic valve, but this option appears to reduce the passivity
of the safety feature.
3-22

5. Independent, detailed evaluation of plant economics and costs
is needed.

6. Thorough seismic analysis is needed to assure the integrity of
the elevated pools and associated piping, valves, and
containment structures.

3.4 LIQUID METAL COOLED REACTORS (LMRs)

3.4.1 Introduction

The three IMR concepts evaluated by NPOVS are the Large Scale
Prototype Breeder (LSPB), the Sodium Advanced Fast Reactor (SAFR), and
the Power Reactor Inherently Safe Module (PRISM). To place the discus-
sion of these designs in perspective, this section begins with a summary
of major design options, design challenges, and design tradeoffs envi-
sioned for commercial ILMRs. This is followed by a description of each
concept and a discussion of their advantages and disadvantages with
regard to the NPQVS criteria. To complete the section, the research and
development needs for the concepts are presented.

3.4.2 Design Options, Challenges, and Tradeoffs

 

Recently, emphasis of the U.S. breeder program has shifted towards
enhanced passive safety, lower capital and operating costs, shorter con-
struction times, and enhanced licensability.® This led to a reexamina-
tion of the many design options, challenges, and tradeoffs which are
available for pursuing the IMR concepts. Those mentioned below are the
more important ones with regard to the NPOVS study.

The universal choice of sodium as the coolant for commercial IMRs
results in several design options and design challenges, all of which
relate to the physical and chemical properties of sodium.’ Design
options made possible by sodium which enhance passive safety include:
(1) a low-pressure primary system, (2) an operating primary coolant
temperature well below boiling, (3) a large heat capacity in the coolant
volume, (4) relatively low coolant velocities and pumping powers, (5) a
coolant which is compatible with preferred cladding materials, and
(6) natural circulation heat transfer loops for decay heat removal.
Design challenges associated with the use of sodium as the reactor
coolant include: (1) protection against sodium fires and sodium-water
reactions, (2) considerations of reactivity effects associated with
loss-of-coolant accidents, (3) minimization of primary coolant activa-
tion by neutron absorption, (4) maintaining the coolant in the 1liquid
state during accident conditions, (5) requirements for nonvisible re-
fueling, and (6) sodium purity monitoring and cleanup. Through the many
years of extensive research, development, and experience with sodium
systems, design tradeoffs have been developed and demonstrated. To a
great extent IMRs are attractive because they use sodium as a coolant;
many design decisions are dominated by considerations of sodium
3-23

properties. However, the U.S. utility industry lacks familiarity with
the use of sodium cooling technology, and this fact must be reckoned
with in securing utility competence.

Another significant option for IMR designs is the required power
output and therefore the size of the reactor core. Presently, a large
core producing about 3500 MW(t) offers plant economy of scale and
relative simplicity of reactor control compared to multiple modules.
Claims associated with producing the same power with several smaller
cores includes construction advantages associated with factory
fabrication, enhanced passive safety against core disruptive accidents,
and greater flexibility in the design of passive decay heat removal sys-
tems. In addition, safety characteristics of these smaller cores can be
demonstrated in full-scale prototypic tests, thereby strengthening
licensing positions. Other claims made for smaller, multiple cores for
each power station include lower investment risk, a better match of com-
pletion schedules with load demands, and high availability.

The NPOVS assessment includes consideration of the entire fuel cy-
cle. A once-through cycle does not take advantage of the potential for
IMRs to breed and thus provide greatly extended fuel reserves. Fuel re-
cycle is anticipated when it becomes cost effective and will be required
for a long~term nuclear capability. Design options which most influence
fuel recycle include long life cores and the fuel type. Oxide fuel has
been the reference for all foreign and domestic programs, but, recently,
metal fuel has been reexamined at the Argonne National Laboratory. See
Appendix E for discussion of fuel cycles for oxide and metal fuels.,
However, other concerns such as safeguards, the availability of fuel
from enrichment or reprocessing facilities, the cost and licensability
of these facilities, and public and utility acceptance must be consid-
ered. Thus, a major business decision for building an LMR may be very
dependent on having an acceptable and available fuel-supply system.

Two design options have traditionally existed for the configuration
of the primary loop in a commercial IMR. The first is a pool design for
which the core, intermediate heat exchanger, and remainder of the pri-
mary system are all contained within a single vessel. The second 1is a
loop design where only the reactor core is placed in the sodium~filled
reactor vessel; the primary pump and intermediate heat exchanger are
outside this vessel, being incorporated into a heat transfer loop. The
choice of a loop or pool option 1is significant since it has a major
influence on the remainder of the plant design. A recent worldwide
emphasis to decrease the capital costs projected for commercial IMRs has
focused attention on the pool design, which generally offers an advan-
tage in compactness.8

3.4.3 Design Descriptions

 

This brief discussion of the LSPB, SAFR, and PRISM designs includes
plant characteristics which support claims concerning economics, safety,
3-24

licensability, constructibility, and public acceptance. Table 3.1 pro-
vides a summary list of the more important features of each design as
presented in the references to this report and updated by the designers.

3.4.3.1 The Large Scale Prototype Breeder (LSPB)

 

This design effort is assisted by several contractors” sponsored by
the U.S. Department of Energy and, since 1982, by the Electric Power
Research Institute.®”12 71t is intended to provide the prototype of a
commercially deployable plant which could produce power before the year
2000. Preliminary design of a four~loop configuration is completed and
provides the basis for our evaluation. Preliminary design of a pool-
type concept has been initiated.

Of the IMR concepts considered here, the 1319-MW(e) LSPB is most
easily associated with historic design evolutions. It incorporates many
features from previous IMFBR designs, both in the United States and
abroad. In addition, the LSPB includes design innovations to reduce
capital and operating costs and enhance passive safety. Thus, this
concept, as presented by the LSPB project, satisfied the NPOVS ground
rules.

As indicated in Table 3.1, each of the four loops of the LSPB con~-
sists of a primary heat transfer loop, an intermediate heat transport
loop, and a steam generator system. The four steam generators feed a
single turbine manifold. The reactor core is heterogeneous using both
Pu02 and UO2 as fuel.

A major design goal of the LSPB has been cost competitiveness.10
The third plant constructed is intended to be economically competitive
with both coal-fired and IWR plants under the assumption of no govern-
ment support. Specific features incorporated to enhance constructibil-~
ity and reduce capital and operating costs include: (1) the capability
to operate at reduced power using three loops while the remaining loop
undergoes maintenance, (2) shallow excavation, (3) close-coupling of
major systems for reduced lengths of large-diameter sodium piping,
(4) reduced sizes of major buildings to reduce capital costs of commod-
ities such as concrete and rebar, (5) use of cable multiplexing to re-
duce cable requirements and eliminate cable spreading rooms, (6) pre-
assembly of subsystems on-site prior to installation in the plant, (7)
separation of the balance-of-plant from the nuclear island, and (8) in-
line arrangement of major buildings to enhance constructibility.
Another significant innovation of the LSPB system is the containment

*The design efforts by General Electric, Rockwell International,
Combustion Engineering, Westinghouse, Burns & Roe, Bechtel and Stone &
Webster are technically integrated by the The Electric Power Research
Institute Consolidated Management Office for the LMFBR.
Table 3.1.

selected for initial investigations

Design and construction characteristics of the IMR designs

 

Design and construction
characteristics

*
LSPB Loop

SAFR

PRISM

 

Power level [MW(e)]

Reactor Exit Tempera-
ture (°C)

Steam Cycle and Steam
Conditions (°C, MPa)

Plant Configuration

Number of Pumps in
Primary Loop(s)

Number of Intermediate
Heat Exchangers (IHXs)
per Power Unit

Number of Intermediate
Loops per Power Unit

1319 [3500 MW(th) and a
net plant efficiency of
37.6%)

510

Benson cycle (454, 15.7),
conventional

Four loops (pool design
is being developed)

Four (one per loop)

Four (one per loop)

Four

Single Power Pak, 350;
Multiple Power Paks,
700, 1050, 1400, etc.

510 (9 Cr—1 Mo used for
entire intermediate
system)

Benson cycle (457, 18.3),
conventional

Pool (each Power Pak has
one reactor and one
turbine)

Two, in primary vessel of
each Power Pak

Four, in primary vessel
of each Power Pak, Pri-
mary flow is gravity
driven on the tube side.

Two, each with its own
Steam Generator, for
each Power Pak

Single reactor, 138; the
smallest power block unit
has three reactors and one
turbine, for 415. Station
power with multiple segments
of 415, 830, or 1245

468

Steam enters the turbine
from a steam drum fed by
three steam generators.
Steam is at saturation
conditions (282, 6.6)

Pool (three pool modules,
each with one reactor, for

each turbine)

Four per module

Four per module

One per reactor module

with one steam generator
per reactor

GCc-¢
Table

3.1. Design and construction characteristics of the LMR designs
selected for initial investigations (continued)

 

Design and construction
characteristics

Fuel

Reactor Shutdown System

Shutdown Heat Removal

LSPB Loop™

U-Pu oxide

Diverse, redundant system
for active shutdown

Passive rod release from
Curie-point or other
temperature effect as a
design option

Normal: BOP using
natural or forced
circulation

Dedicated: Two, inde-
pendent, diverse, and
redundant safety grade
systems that remove heat
directly from the reactor
vessel to the atmosphere.
One system uses natural
circulation

Backup: Cross-connection
to the natural circula-
tion heat removal éystem
of the exvessel, fuel
storage tanks

SAFR

U-Pu oxide, or U-Zr
metal, or U-Pu-Zr
metal

Diverse, redundant system
for active shutdown

Self-activated, passive,
temperature~induced
release of these rods
from Curie-point effect

Normal: BOP with natural
circulation.

Investment Protection:
Direct reactor auxiliary
cooling (DRAC). Sodium-
to—air heat transfer us-
ing natural circulation.
Safety: Reactor air
cooling system (RACS).
Reactor guard vessel

is cooled by natural
circulation of air

PRISM

U-Pu oxide or U-Pu-Zr metal

Diverse, redundant system
for active system

Passive shutdown from
negative reactivity feed-
back due to temperature
increases and self actua-
ted release of shutdown
rods from over—-temperature

Normal: Heat transport
system to turbine—-generator
condenser

Investment Protection: Air
cooling of steam generator
shell.

Safety: Radiant Vessel
Auxiliary Cooling System
(RVACS). A passive, radiant
heat transfer, natural
circulation system that
operates efficiently at

high temperatures

9¢-¢
Table 3. 1.

selected for initial investigations (continued)

Design and construction characteristics of the LMR designs

 

Design and construction
characteristics

IHTS and BOP Configuration

Fuel Cycle Facilities
and Strategy

Construction Character-
istics of Major Compo-
nents and Structures

Containment/Confinement
Building Characteristics

Net Thermal Efficiency (%)

Availability (%)

LSPB Loop™

Modularization is
stressed. A rectangular
reactor containment
building will be used.
Systems outside contain~
ment are of conventional
design

Under-the—head reactor
refueling. Assumes off-
site fuel reprocessing.

Modular construction
will be emphasized.
Access plugs are pro—-
vided in the top of the
containment building
for removal of large
components.

Reactor, PHTS, and
auxiliary equipment in
a concrete containment
building enclosed by

a steel confinement
structure.

37.6

80 (design)

SAFR

IHTS loops and components
made of 9 Cr—1 Mo steel.
Conventional and non-
safety grade, seismic II
design.

On—-site reprocessing and
refabrication included in
design layout but little
detail design as yet.
Spent fuel is stored for
a year in the primary
vessel.

Reactor assembly with
vessels, internals and
deck is factory built

and barge shippable. Ac-
cess plugs provided in
the top of the contain-
ment building for removal
of large components,

One containment building
for each Power Pak.

36.7

>84 for a single Power
Pak.

PRISM

High commercial grade to
reduce cost. The reactor
module and refueling equip-
ment are of nuclear safety
grade.

Refueling using a mobile
refueling machine which
moves from one module to
the next. Reprocessing
and refabrication may take
place either on—- or off-
site.

Reactor modules are shop
fabricated and assembled and
are rail shippablie. In ad-
dition, the intermediate
sodium loop, the steam
generator, and other BOP
systems will be modularized
and factory produced.

Each module has its own
containment.

32.5

88 estimated; 80 used in
economic assessments,

LT-¢
Table 3.1.

selected for initial investigations (continued)

Design and construction characteristics of the IMR designs

 

Design and construction

 

characteristics LSPB Loop™ SAFR PRISM
Plant Lifetime (yrs) 40 60 40 (but reactor modules
can be replaced at
relative low cost)
Core Design Characteristics
a) Type Heterogeneous Heterogeneous Homogeneous
b) Height (meters) 4.83 3.25 1.76
¢) Diameter (meters) 5.71 4,04 1.93
d) Resident time in 3 (4 for advanced core) 4 4
core (yrs)
e) Refueling Intervals 1 1 1
(yrs)
f) Cover gas Ar He He
Reactor Vessel
a) 1ID (meters) 14.6 11.9 5.8 {(containment)
b) Height (meters) 19 14.5 19.5
Burnup (MWd/kg) 109 158 107

Breeding Capability

Doubling time of 25 years
for breeder core reload.

Breeding is not required

for the initial core.

System will need only a
feed of U-238 since
Pu~239 needs will be
supplied by conversion.

Breeding ratio of 1.04
for oxide fuel and 1.22
for metal fuel.

 

*The LSPB pool concept has similar characteristics but offers a higher power level and plant efficiency
It also will require a larger vessel (19 m ID, 21 m H).

[1350 MW(e) and 38.5%] and improved shutdown heat removal.

8¢t
3-29

design which is a rectangular, steel-lined, concrete building with roof
hatches for construction and maintenance. Adjacent nuclear island
buildings are integral with the containment thereby providing cost-
effective containment and confinement capabilities. The extent to which
the LSPB design has achieved lower capital costs is suggested by
comparison to the CRBRP design. The LSPB plant, while producing four
times the net electrical power of the CRBRP design, occupies a nuclear
island which is physically smaller than that of CRBRP. Finally an
option has been maintained to use a fuel designed for cost performance
by reducing the breeding specification. These design studies indicated
that fuel cycle costs could be reduced by about 3 mills/kWh.

Design modifications under consideration could enhance inherent
protection for failure-to-scram events through temperature-induced ex-
pansion of control rod drive-lines or temperature-induced control rod
releases or other Self Actuated Safe Shutdown (SASS) type devices.
Calculations are being conducted to identify the design measures needed
to assure no boiling for a loss of flow with trip failure. 1In addition,
the LSPB decay heat removal capability is enhanced by incorporating the
capability for natural circulation in the normal heat transport sys-
tems. These design features should increase the licensability and ac-
ceptance of the plant by the public and the utilities.

Additional design features associated with safety and licensability
include the use of two independent and diverse reactor shutdown systems
and the use of two independent and diverse, safety-grade, decay heat
removal systems. One of these decay heat removal systems consists of
two, forced-circulation loops and the other is a passive, natural circ-
ulation loop. Both decay heat removal systems use the outside air as
their ultimate heat sink and sodium in the reactor vessel as the heat
source. The LSPB also utilizes a heterogeneous core design. Because of
these enhanced safety features, the LSPB balance of plant (BOP) design
has been downgraded from safety grade to commercial codes to obtain cost
reductions and enhance constructibility.

3e4.3.2 Sodium Advanced Fast Reactor (SAFR)

 

This plant, being designed by the team of Rockwell International,
Bechtel, and Combustion Engineering, for the U.S. Department of Energy,
consists of one or more independent power generating units called Power
Paks, as illustrated in Figure 3.5.13-16 The utilization of multiple
units at one site permits cost savings through sharing certain facili-
ties and services. These shared facilities include the control build-
ing, the plant service building, the nuclear island maintenance build-
ing, and the fuel cycle facility if colocated with the power plant. A
major goal for the initial SAFR design effort was to determine the Power
Pak power level, and therefore size, which is the optimum trade-—-off of
cost, passive safety, utility acceptance, licensability, and constructi-
bility. Factors which influenced the selection of the 350-MW(e) size
included short construction times, low investment risk, economy of
scale, and moderate energy costs. For the basic design configuration of
each Power Pak, Rockwell made effective use of their previous IMR design
3-30

ORNL—-DWG 86—4051 ETD

 

 

 

 

 

 

 

 

 

 

 

 

Rockwell Inter-

Source:
"SAFR Discussions at ORNL,

Power pak elevation for SAFR.

Fig! 3!5!
national,
1985."

 

January 11,

Rocketdyne Division,
3-31

experience, particularly that associated with the Large Pool Plant
(LPP).17 Advanced LMR technology and enhanced passive safety features
introduced into the design are listed as follows: (1) metal fuel and
its associated reprocessing innovations have been retained as an option,
(2) redundant and passive decay heat removal systems have been employed,
(3) a relatively high primary system temperature was selected with the
use of an advanced material, 9 Cr—l Mo, for the entire intermediate
loop, (4) a backup, self-actuated shutdown system has been included, and
(5) heterogeneous core designs with self-regulating characteristics have
been incorporated with the objectives of limiting the potential effect
of hypothetical accidents.

As indicated in Table 3.1, each 350-MW(e) Power Pak consists of a
reactor vessel, primary and intermediate heat transport systems, a steam
generator system, and a turbine generator. Safety-related systems and
components are minimized and localized in the design such that nuclear
safety is decoupled from the BOP and Intermediate Heat Transfer System
(IHTS). The reactor assembly is factory built and barge shippable. It
contains the primary system and a spent—fuel storage rack. Fuel trans-
fer is by a hoist mechanism and rotating plug which is part of the ves-
sel head closure. Included in the primary system are the reactor, two
inducer-type primary pumps, and four intermediate heat exchangers
(IHXs). In each of the two independent, intermediate loops, non-radio-
active sodium is circulated through the IHXs and a booster-tube, hockey-
stick steam generator operating in the once-through mode. The super-
heated steam from the two steam generators (one for each loop) is
directed to the turbine generator. The reactor containment building for
each Power Pak encloses the reactor vessel and the in-containment,
conventional (A-frame) fuel handling system. This building is a rec-
tangular, reinforced concrete structure with a flat roof. Hatches are
provided in the roof to facilitate handling of components, if necessary,
thus limiting the building size and hence the coanstruction commodities
required. The reactor guard vessel constitutes part of the containment
envelope. The non-safety-grade, steam generator building for each Power
Pak is a conventional building mounted on the base mat.

The normal mode of decay heat removal uses natural circulation of
sodium through the heat transport systems of the Power Pak. 1In addi-
tion, two independent, natural circulation, backup systems are pro-
vided. The first is a direct reactor auxiliary cooling system (DRACS)
which transfers heat from the primary pool to the outside air using a
sodium-to—air heat exchanger. The second is a passive, safety-related,
reactor air cooling system (RACS) which operates with natural circula-
tion to provide the ultimate decay heat removal capability through cool-
ing of the reactor guard vessel. The RACS also provides passive cooling
of the reactor cavity. The diverse and redundant shutdown system con-
sists of both primary and secondary control rods as well as a self-
actuated inherent shutdown system which responds to sodium overtempera-
tures.

The site construction time for a Power Pak unit, from ground
breaking to initial power operation, 1is estimated to be thirty-three
months., The 1licensing plan for SAFR stresses standardization and a
3-32

prelicensed Power Pak so that only site-related licensing considerations
are required after obtaining a Final Design Approval.

3.4.3.3 Power Reactor-Inherently Safe Module (PRISM)

 

The PRISM concept of General Electric is being designed under con-
tract for the U.S. Department of Energy.l8 21 A gsimplified drawing of
the concept is shown in Figure 3.6. A major design emphasis of the
PRISM concept is incorporation of passive safety through use of: (1) a
relatively low power reactor core of 133 MW(e), (2) a pool design with
relatively low primary sodium temperatures, (3) a safety-grade passive
decay heat removal system, and (4) large negative temperature reactivity
feedback in the core design with the iatent of limiting potential core
disruptive accidents to the initiating stage. Another major emphasis of
the PRISM concept is licensing by demonstration of plant safety through
tests conducted with at least the primary system of a prototype reactor
module at a test facility.

This safety-grade reactor module is the basic power-producing unit
in the PRISM design. The low-pressure, primary system of each module is
a pool-type design with the reactor core, four cartridge-type, electro-
magnetic primary pumps, and four cartridge-type intermediate heat ex-—
changers all contained within the reactor vessel. The intermediate sys-
tem associated with each module consists of a single loop which trans-
fers heat energy from a common header, fed by the four intermediate heat
exchangers, to a single steam generator. Thus, the primary loops and
the single, intermediate loop associated with each reactor module are
independent of those of other modules.

The common tie between the reactor modules occurs on the turbine
side of the steam drums. Steam from three steam generators drives a
single turbine. Therefore, the PRISM design, 1like the HTR concept
considered by NPOVS, has multiple reactors and their associated heat
transport system supplying steam to a single turbine.

This power unit, or segment, containing three reactors, three steam
generators and one turbine produces about 415-MW(e) of power. A power
station, in turn, would consist of one or more of these segments. The
reference PRISM design produces 1245-MW(e) and has three segments for a
total of three turbines and nine reactors. FEach segment is functionally
independent of the others.

The homogeneous reactor core is fueled with U-Pu oxide. Through-
the-head refueling will occur once each year using a mobile refueling
machine. The residence time of the fuel is 4 years. The radial blank-
ets containing UO2 contribute to a breeding ratio of about 1.03, de-
signed to compensate for losses during recycle. The diverse and redun-
dant control and shutdown system contains six primary control rods and
two secondary control rods.
3-33

ORNL—DWG 86--4052 ETD

— RVYACS AIR OUTLET

-

 

ey

RYACS AIR
INLET

 

 

 

 

 

 

 

 

 

 

 

  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

- ". | _~CONTAINMENT VESSEL
HODULAR THX (8| :/(//,///-(1910 x 54'H)
v s
10 | ,~.///,»{M PUMP (4)
t : |
; | [/;/
ol ] A
- ! |  R
. .L'
1 ‘ .
M
|

 

Fig. 3.6. The below grade modular concept for PRISM, October 1984.
Source: General Electric Company.
3-34

The containment vessel is 5.79 meters in diameter and 19.5 meters
high. The vessel is shop fabricated and assembled and rail shippable.
It is installed below grade to facilitate ground-level refueling, to re-
duce building costs, and to provide a natural barrier to missiles. A
sodium containment vessel surrounds the reactor vessel and is sized so
that the reactor core will always remain covered by sodium even if the
reactor vessel should develop a leak. Details of the containment/con-
finement design are still under consideration. The primary pumps and
intermediate heat exchangers can be removed easily through the top head
for maintenance.

The reactor vessel and containment vessel are important components
of the safety grade, shutdown heat removal system. Normally this resid-
ual heat would be removed by the non-safety grade, secondary heat trans-
port loop associated with each reactor module. If this normal heat path
is not available, the safety grade Reactor Vessel Auxiliary Cooling Sys-
tem (RVACS) would provide this safety function. The RVACS is a passive,
natural circulation system that is always in operation. Radiative heat
is transferred from the reactor vessel to the containment vessel. This
heat is removed to the atmosphere by natural circulation of outside air
past the outside surface of the containment vessel. Calculations by GE
indicate that this system can accommodate decay heat removal require-
ments after loss of normal heat removal capability concurrent with a
reactor scram. For this case, the peak in the primary sodium tempera-
ture would be about 600°C and would occur several hours after the start
of the event. An important aspect of the RVACS system is that its heat
removal capability increases substantially with increasing primary
sodium temperature.

The passive safety features of PRISM are further indicated by its
response to the very severe and unlikely accident where the loss of
primary coolant pumping power, the loss of normal heat sink, and a fail-
ure to scram all occur at the same time. The GE analysis of this hypo-
thetical event with no operator intervention predicted that, after some
initial oscillations in core reactivity and temperature, an equilibrium
situation would be reached within about ten hours without exceeding
allowable temperatures. At this equilibrium state, the heat generation
rate of the critical core would be matched by the heat rejection rate of
the RVACS system with a system temperature of about 630°C.

Factory fabrication and assembly, standardization, and a reduction
in systems required to be safety grade have been stressed in the PRISM
design as a means of offsetting a perceived diseconomy of scale for
small units. Advantages projected for this construction technique
include more efficient use of site labor, a much shorter construction
time of three years from start of construction to full power operation,
"learning curve” benefits due to replication, and a closer potential
match of a utility's power production capabilities to its load. Since
the reactor module is the only nuclear qualified component, the balance
of plant can be constructed economically to high quality industrial
standards.
3-35

The PRISM licensing plan calls for prelicensing of a prototypic
reactor module so that only site specific issues need be addressed for
licensing a plant. This prelicensing would be accomplished through a
design and safety test program during which the basic safety and
economic claims for the concept would be demonstrated by prototypic,
full scale tests.

3.4.4 Advantages and Disadvantages of the IMR Concepts
with Regard to the NPOVS Criteria and Essential
Characteristics

 

 

 

3e4e4.1 General Overviews

 

Commercialization and marketing of an IMR in the anticipated market
between now and around the year 2010 may be difficult to accomplish.
Not only do IMRs have the same negative market factors as other con-
cepts, including an uncertainty in the need for power, licensing chal-
lenges, and financial uncertainties, but IMRs must also overcome addi-
tional concerns such as higher capital costs associated with traditional
designs, their perceived role only as breeders, a lack of wutility
experience with IMRs, and uncertainties associated with an adequate and
cost competitive fuel cycle. In fact, one could argue that IMRs will
penetrate this market only if they have a unique and very important
advantage over other power generating concepts.

Such an advantage may arise from the innovative ILMR designs evalu-
ated here. Their strong emphasis on cost reduction, passive safety,
rapid construction, licensability, and low economic risk are certainly
appropriate to meet the challenges of future markets. 1In the discussion
which follows, the advantages and disadvantages, or challenges, outlined
above will be discussed in the same order as the NPOVS criteria, essen-
tial characteristics, and desirable characteristics presented in Section
2+.2.1+ Many of these comments apply to all of the IMR concepts and they
will be presented first. These will be followed by comments specific to
a particular concept.

3.4.4.2 Advantages of the IMR Concepts

 

1. Public Risk: A significant feature of IMRs is the passive
safety which may be incorporated into their designs.7 Among
the passive features 1is the tendency for sodium to provide
natural convection cooling, the high thermal conductivity of
sodium, the large heat capacity of the reactor system (which
affords long grace periods for problem diagnosis and correc-—
tive action), low-pressure design, and operating temperatures
far below the boiling point of sodium. The usefulness and ef-
fectiveness of these features were successfully demonstrated
in tests at several plants including the Prototype Fast Reac-
tor (PFR), Phenix, and the FFTF. They are utilized in all
three of the designs considered here. Because of passive
3-36

safety features, these designs require fewer engineered
(active) safety systems and less emergency power than conven-
tional IWRs. One caution is that, since these designs rely on
outside air as the final heat transfer medium/sink for decay
heat removal, they might be susceptible to common cause
external degradation events, such as fires or dust storms .22

The three designs considered here incorporate a reactor
shutdown system similar to the CRBRP concept. They may en-
hance the passive safety of their system with respect to Hypo-
thetical Core Disruptive Accidents (HCDAs) through a passive
control-rod release mechanism which will be activated by high
sodium temperature. This feature may terminate any over-heat-
ing event before sodium boiling occurs. The designs also limit
the total amount and rate of reactivity insertion possible in
the event of a control-rod withdrawal accident. In additionm,
the assurance of decay heat removal capabilities is provided
by both active and passive systems which incorporate signif-
icant redundancy and diversity. Finally, for the PRISM and
SAFR designs in particular, the core and control drive Ilines
are being designed so that many HCDA initiating events will be
terminated by feedback responses from temperature increases
and resulting thermal expansion before core degradation
initiates.

In our judgment, IMR designs can meet and probably sig-
nificantly exceed the goals of NPOVS Criterion 1. For exam-—
ple, the PRA study completed for the CRBRP calculated a core
damage frequency for an HCDA to be 3.6 x 10~ °/year, with
seismic events being the major initiator.23 The frequency for
internal initiators was about a factor of ten less. Another
independent study for the SNR-300 plant in Germany concluded
that, "both the frequency of major accidents and the extent of
damage associated with such accidents are smaller than those
estimated in the German Risk Study for the PWR-1300."2% These
designs achieve 1low HCDA probabilities to a great extent
because of the reliability of active safety systems, partic-
ularly the diverse and redundant reactor shutdown systems.
Credit for inherent or passive responses of the core which
could result 1in early termination of the event are incor-
porated into the calculations in a conservative manner.

Investment Risk: In our judgment, the IMR designs can meet
and probably significantly exceed the goal of NPOVS criterion
2. The emphasis on simplicity of design, the use of fewer
complex safety systems, and the incorporation of passive
design features, discussed under criterion 1, would all
contribute to low investment risk. In addition, extensive
reliability studies and PRA evolutions are planned for each
design.
3-37

Economic Competitiveness: The capability of breeding signifi-
cantly more fuel than is consumed in producing power is a
major long-term advantage of IMRs. This breeding capability,
coupled with a complete fuel cycle, would enable LMRs to ex-
tract between 60 and 80 times the energy from a given quantity
of natural uranium than can be done using non-breeders.’ In
addition, IMR operating costs need not be as sensitive to fuel
costs as non~breeders. The IMR designs can offer breeding as
a design option to be implemented by a relatively easy and in-
expensive core modification when it becomes economically
attractive to do so. Comparative evaluations reported in
Chapter 3, Volume III, of this report indicate a potential
competitiveness with both the best IWR experience and with
coal-fired plants.

The LSPB concept has perceived economy-of-scale advan-
tages and has incorporated significant cost reduction features
and a short construction schedule into the design. The abil-
ity to add plants in smaller power increments, thereby better
matching utility needs, is a potential advantage of the PRISM
and SAFR designs. Their lower capital risk achieved by
modular construction and very short construction times is also
attractive. However, it is not clear how costs for the fac-
tories to build these modules will be assessed and costs for
the fuel cycle will be incorporated. This may increase the
cost of the first several plants, and it may be difficult to
justify the high initial costs for factory automation which
would improve manufacturing efficiency. SAFR plans are to use
existing facilities with increased automation for vessel
assembly production up to a few units per year.

Probability of Cost/Schedule Overruns: All three concepts
have stressed constructibility and simplicity, and a complete
design before construction. They utilized modular construc-
tion of major components im a factory and shipment to the
site, and non-safety grade construction at the site for the
BOP. These approaches should minimize delays and cost over-—
runs attributable to quality assurance problems and large con-
struction crews. There is a lack of U.S. industry experience
in IMR construction. However, recent documentation of con-
struction experience indicates that construction problems are
more a function of the management and construction team and
their interaction with the NRC than the reactor type.23 The
concept of learning by experience should apply to SAFR and
PRISM if additional modules and Power Paks can be built by the
same team without delay after completion of the first plant
segment. This can be done while the first segment is pro-
ducing power, but care must be taken to avoid jeopardy to the
operating unit by the construction activities where close
proximity is required such as in the control building.
6.

3-38

Licensability. Assurance of licensability before construction
is emphasized by these IMR commercialization plans. Each
stresses early approval by NRC of a standard plant design.
Thus, only site-specific NRC concerns would need to be
addressed for licensing of subsequent plants. The 1licens-
ability of the LSPB should be relatively high because the
design basis accident analysis and many key safety design
features are based on the CRBRP licensing experience.

The first choice for PRISM licensing, and an alternative
for SAFR, calls for demonstration of the plant's passive pro-
tection against traditional HCDA initiators through tests of a
prototypic reactor module. This concept of licensing by test
has attractive features. Chief among these are a possible re-
duction of analyses, validation of computer codes, and demon-
stration of safety claims to the public, potential investors,
and the NRC.

Some precedence has been established for such tests
through the extensive program at the Southwest Experimental
Fast Oxide Reactor (SEFOR) which demonstrated the effect of
the Doppler coefficient on power excursion,l!? and the recent
tests at Raposdiel® and EBR-II where loss-of-flow HCDAs were
initiated and subsequently terminated by passive feedback of
the core.

The SAFR designers indicate that a possibly more cost ef-
fective approach involves resolving the main licensing issues
by extrapolation of test results from FFIF and EBR-1I. Then a
plant installed on a utility grid would be the vehicle for ob-
taining a standard plant FDA with rulemaking to apply to sub-
sequent plants of the same design.

Demonstration of Readiness: Furope and Japan, which have less
abundant natural supplies of fissile material, perceive a need
for breeders sooner than the United States. For this reason
these countries are pursuing a vigorous program of demonstra-
tion and commercialization of the entire LMFBR fuel cycle. One
can estimate from projects now in place that 50 plant-years of
operation could be compiled by IMR demonstration plants by the
year 2000.8 From past experience, acceptable performance is
expected from these plants. For example, since 1973 the
French, 250-MW(e) Phenix prototype plant has operated with an
overall capacity factor of 60%.8 This experience base will be
relevant to the requirement for a successful demonstration
plant.

A strict interpretation of Criterion 6 requires that
demonstration plants for the specific ILMR plant concepts be
built and operated in the United States before a utility
decision to buy 1is made. To accomplish this task within the
NPOVS time frame is a significant challenge. Nevertheless,
3.4.4.3

1.

3-39

our judgment from evaluations of the marketing and commercial-
ization plans for LSPB, SAFR, and PRISM is that implementation
of any of these plans with a dedicated effort could result in
satisfying this criterion.

Owner Competence: There are many similarities in the oper-
ation of LMRs and LWRs, particularly with regard to reactor
control and BOP functioning. Thus a significant fraction of
LWR operator training and experience would be relevant to
LMRs. In addition, worldwide experience indicates that LMRs
are relatively easy to operate and maintain. Personnel spe-
cifically trained in the operation of sodium systems within
the United States are at national laboratories, industrial
test facilities, and at the U.S. operating IMRs, EBR-II and
FFTF.

Essential Characteristics: The IMR concept designers have
stressed shop fabrication, minimizing nuclear grade compo-
nents, standardization, long plant lifetime, ease of construc-
tion, and passive safety features. The PRISM and SAFR designs
offer a variety of plant sizes to match load growth and, as
explained in Chapter 3 of Volume III dealing with economics,
some availability advantages may result from smaller, multiple
reactor cores.

Desirable Characteristics: Relatively high thermal efficien-
cies (=40%) have been achieved with IMR designs and very low
radiation exposures to workers (on the order of a few man-rems
per year) have been experienced in demonstration plants. En-
hanced diversion and proliferation resistance is possible with
on-site fuel recycle and with the metal fuel option. Fuel
elements can be retained in the core for several years, there-
by yielding burnup values >100 MWd/kg.

Disadvantages of the IMR Concepts

 

Public Risk: Unlike IWRs which are designed to maximize k. ¢¢,
an ILMR under normal operating conditions is not in its most
reactive configuration. Thus, loss of sodium coolant from the
core or core compaction could result in a reactivity increase.
The designs considered here provide protection against loss of
sodium inventory due to leaks and have substantial mitigating
features — which are amenable to demonstration — for accommo-
dating hypothetical accidents even beyond the design basis.
Nevertheless, the way in which traditional licensing concerns
associated with hypothetical accidents are addressed will need
to be fully developed.

Investment Risk: In addition to the comments made under
public risk, some concern still exists about the performance
and reliability of IMR steam generators. Data which could
verify the performance of current designs should be available
3-40

within the NPOVS time period from component testing programs
and foreign plant experience.

Economic Competitiveness: Evaluations of prototype LMR
designs and foreign construction experience indicates that the
capital costs for LMR commercial plants, based on traditional
designs of the 1980s, could be substantially higher than
present LWRs. This higher capital cost, resulting in part
from the need for an intermediate loop, could be compensated
by lower fuel costs and higher efficiencies for LMRs. Higher
efficiencies for LMRs have indeed been realized; the Phenix
plant, for example, has a gross efficiency of 44%.l1%® But, as
indicated below, it is not clear that the potential fuel-cycle
cost advantage for LMRs will be realized within the NPOVS time
constraints. Longer core 1lifetimes are being studied in
future plans. In summary, cost competitiveness can not be
claimed for operating LMR demonstration plants and, assuming
no dramatic changes in fuel costs within the NPOVS time frame,
competitiveness of commercial LMR plaunts can best be achieved
by significant reductions in capital costs.

Probability of Cost/Schedule Overruns: No specific disadvan-
tage identified except that these are new design concepts with
no direct base of experience.

Licensability: The merits of licensing by prototypic tests
have been discussed previously. There are, however, some lim-
itations of this approach. Not all safety claims or hypothet-
ical accident sequences can be demonstrated, and analysis of
accident sequences may still be required. 1In addition, this
could be an expensive test program even if the module can
subsequently be used commercially since the test program could
last several years and analyses of pre— and post-test results
could be a significant effort. On the other hand, the PRISM
designers believe this demonstration to be relatively 1less
expensive for a small reactor when compared to the potential
costs and risks associated with licensing a large reactor.

An alternative would be to use the demonstration facility
not only as a test of the PRISM and/or SAFR designs but also
as an advanced research and development facility for general
LMR passive safety features tests. It could demonstrate reac-
tivity feedback effects as well as provide data for code
verification. Perhaps alternate cores, metal and/or carbide,
could be designed for the same facility. Passive shutdown
systems and decay heat removal systems could be demonstrated
as well. However, its utility for some of these purposes
should be evaluated with respect to the FFTF and EBR-II
capabilities.,

In addition to licensing by test, other LMR 1licensing
igssues would still need to be considered for the standard
plant designs. Prominent among these issues will be the
3-41

site-suitability source term, safety functions and design
decisions associated with containment, passive features which
accommodate HCDA concerns, and the need for redundancy and/or
diversity within and in addition to safety systems which are
passive.

Although wuseful experience was gained through FFTF and
CRBRP interactions and 1licensing activities with the NRC,
licensing rules, guidelines, and procedures are not as well
established for LMRs as for LWRs. However, preliminary
discussions have been initiated with NRC for the IMR concepts.

Demonstration of Readiness: Providing funding for an IMR
demonstration plant will be a significant challenge.

Owner Competence: Even though a large number of utilities
participated to varying degrees in the CRBRP, experience in
LMR operation does not currently exist within the U.S. utility
organizations, and the FFTF and EBR II afford only part of the
requirement.

Perhaps a more pertinent question is whether the owner/
operator could be convinced to purchase a new reactor concept
for which utility experience is limited. This latter need is
perhaps most clearly evident when one considers aspects of the
LMR fuel cycle. 1In short, each fuel cycle option appears to
have some significant difficulties. To provide unique IMR
advantages associated with breeding, such as relative freedom
from concerns about a reliable fuel supply, a complete fuel
cycle should be utilized. This means that proven and reliable
on-site or off-site reprocessing, refabrication, and waste
handling of suitable scale mst be available to the owner/
operator at a reasonable cost. The basic technology required
for IMR fuel cycles has been developed in the United States
and demonstrated overseas, and the first few LMRs could be
supported by small-scale development facilities. However, if
one assumes that this capability will be provided on-site,
then uncertainties associated with available trained person-
nel, cost, safeguards, reliability, licensability, and public
and utility acceptance are envisioned. (See also Appendix E).

It is not difficult to conclude, for example, that costs
savings or other incentives must be significant and proven by
experience before a wutility would choose to purchase and
operate a reprocessing plant. Technical and organizational
options making this concept more attractive include a less
complex fuel cycle, the IFR concept for example,26 or the
option that some other institution (not the utility) operate
all facilities except (or including) the power station.
These, and perhaps other options, could improve the viability,
3-42

but acceptance of this concept by a utility and its implemen-
tation and demonstration in the NPOVS time frame seems
unlikely.

1f, on the other hand, one assumes that off-site, cen-
tral, reprocessing facilities would be used to complete the
fuel cycle, it is difficult to envision the economic need for
commercial facilities of this type much before the middle of
the 21st century. Thus, off-site reprocessing may not be
available in the United States within the NPOVS time frame.

Still another option for closing the fuel cycle is to
rely on other countries to provide this service. Difficulties
associated with this option include problems associated with
Pu shipments between countries, adverse balance of payments,
and the assumption that such a commercial industry will in
fact be available to the United States.

If counting on a commercial fuel reprocessing industry is
imprudent, another option is to consider a once-through cycle,
including the possibility of spent fuel storage until commer-
cial reprocessing/refabrication facilities are available.
Difficulties associated with this choice are economic (tradi-
tional IMRs with once-through fuel cycles would have fuel
costs about twice those with Plutonium recyc1e27) and institu-
tional. Once—-through cycles may need to use 235y enriched to
20 to 30% which are levels beyond present production for com—
mercial use. The once~through option could likely be enhanced
by the incorporation of low-power density, heterogeneous,
long-lived (10 years or more) core designs.

Essential Characteristics: Maintenance requirements and oper-
ating staffs for PRISM, and to a lesser extent SAFR, may ex-
ceed those for plants with a single reactor. Security staff
requirements for PRISM can be small because of underground lo-
cation and inaccessibility of key safety features during oper-
ation. On the otherhand, regularly scheduled refuelling and
maintenance reduces the need for extra manpower peaks at an-—
nual refuelling in a monolithic plant. 1In addition, design of
the control system for PRISM must accommodate multiple reactor
cores providing the main source of energy to a single tur-
bine. Licensing requirements, particularly those associated
with the option of reprocessing and refabrication of fuel on-
site, are not completely defined. 1If the overall nuclear in-
dustry, including government support, continues to decline,
the availability of qualified vendors may be in question.

Desirable Characteristics: On-line refueling, though consid-
ered, has not been incorporated into any designs. The PRISM
plant, however, does have the capability to generate electric
power continuously while a single module is being refueled.
Completion of the fuel cycle, important for freedom from fuel
3-43

supply concerns and accomplished in foreign programs, has not
been accepted in the United States because of economic and in-
stitutional considerations.

3.4.5 Research and Development Needs for the LMR Concepts

 

3.4.5.1 Introduction

Two different perspectives are presented in connection with IMR re-
search and development (R&D) needs. First, the viewpoint of the plant
designer is reflected through a collation of design—specific R&D re-
quirements for the three LMR concepts considered in this report. Then
consideration 1s given to general goals for the U.S. IMR R&D program
which could contribute to a healthy and competitive industry considering
the worldwide marketplace.

3.4.5.2 Design—-Specific R&D Requirements

 

Each designer of the three LMR concepts considered by NPOVS recent-
ly completed an assessment of specific R&D needs and reported conclu-
sions, 2830 Appendix D presents summaries of these needs, where in sev-
eral instances, similar needs have been combined. These design-specific
needs can be classified as follows: (1) advanced core design tasks
which include developing improved neutron counting channels, evaluating
shielding designs, testing self-actuated shutdown systems, performing
PRA assessments, and evaluating responses to accidents; (2) shutdown
heat removal experiments and analyses to evaluate design effectiveness,
design margins, and immunity to external events; (3) fuel related activ-
ities such as evaluations of metal fuel cycles, high burnup tests of
oxide fuels, and performance of these two fuels during upsets or when
breached; and (4) system— and component-related studies emphasizing
operating plant experience, scale model flow and temperature tests, in-
corporation of advanced instrumentation and control technologies, and
improving steam generator performance.

A large base of test experience exists for the oxide fuel but that
for metal fuel is limited. It is anticipated that an extensive fuel
testing program would be required for metal fuel before proceeding to
commercial use. 1In the French qualification of oxide fuels for LMFBR
use, the testing program included an extended operation with refabri-
cated fuel from the reprocessing demonstration. A similar effort for
metal fuel may be prudent. Reprocessing and refabrication are discussed
more extensively in Appendix E.

3.4.5.3 General R&D Goals for the U.S. National IMR Program

 

A necessary but perhaps not sufficient list of goals for IMR R&D
includes the following:

1. Develop an IMR design which has a clear, unique, and significant
advantage in the marketplace over other concepts. The current de-
sign studies are judged to be consistent with this goal. However,
2.

3-44

a small or debatable advantage for IMRs may not be adequate for
penetration of a market dominated by LWR designs. Present programs
are appropriately directed toward the innovative design of a cost
competitive, modern (i.e., incorporating new technologies), and in-
herently safe IMR. Licensability advantages as well as public and
utility acceptance also are important reasons for this goal to be
achieved.

Maintain the option for rapid incorporation of breeders and of a
complete fuel cycle into the future marketplace. The potential
long-term market for breeders is assured unless nuclear fission
energy 1is to have only a transitory role. Also, there exists a
possibility for substantially increased shorter term demand if, for
example, increased burning of coal should be found unacceptable.

Complement the IMR R&D being performed by Europe and Japan so that
the United States will be in a strong negotiating position to ex-
change our accomplishments for experience from their more acceler-
ated programs of demonstration and commercialization. Programs
which typify contributions to this goal include advanced computer
code development, materials research, licensing reform, advanced
designs, metal fuel research, advanced instrumentation, contrel and
simulation, and development of double-wall steam generators.

Contribute to a reduction in licensing concerns, costly design mar-
gins, and special systems resulting from the potential for core-
disruptive accidents. Each of the IMR designs counsidered by NPOVS
have already contributed to this goal. Advances in the future
should stress demonstration of passive safety features, computer
code validation, and experimental verification of specific reac-—
tivity feedback effects incorporated into designs.

Demonstrate, test, and utilize to the fullest extent possible ad-
vanced technologies, components, and design concepts. Implementa-
tion of R&D to satisfy this goal would increase the available de-
sign options,31 thereby increasing the likelihood of optimizing the
design to accomplish a larger number of desirable objectives and
specifications. These advanced technologies could include automa-
tion, research resulting in higher plant operating temperatures and
efficiencies, use of artificial intelligence, and increased use of
computers for control and simulation, surveillance and diagnostics,
data display and verification, and maintenance functions. Automa-
tion may be very important to the licensing and economic operation
of multiple modules which feed a common steam system.

Study and develop containment concepts which both simplify the
overall nuclear system and ensure protection against both internal
and external events, which may be judged credible. This work must
be coupled closely with source term evaluations.
3-45

7. Investigate IMR core designs which might be competitive using a
once—-through fuel cycle. These studies should include the poten-
tial use of Pu obtained from foreign sources. This task will prob-
ably require determining an optimum core geometry, power density,
core lifetime, and neutron energy. It could contribute signif-
icantly to competitive LMRs for a scenario of low energy-growth-
rates. One such concept is an ultra long-life core which would
require refueling only at major inspection intervals of approxi-
mately every ten years.

8. Develop and demonstrate technical solutions to the challenges asso-
ciated with the IMR fuel cycle which were identified in the
previous section of this report.

3.5. HIGH TEMPERATURE REACTORS (HTRs)

The focus of NPOVS HTR evaluation is on the modular HTR with the
steam generator and core in separate steel vessels connected by con-
centric crossducts in a side-by-side configuration. An extensive amount
of information has been derived from the DOE HTR Program.”"‘38 To place
the safety and economic features of the modular HTR in perspective, the
large HTGR [2240 MW(t), 860 MW(e)], which was the focus of the DOE Pro-
gram for several years, is carried by NPOVS as a point of reference. A
summary of its advantages, disadvantages, and R&D needs can be found in
Refs. 39 and 40 respectively. Appendix F presents the general design
features of a large HTGR as a reference for HTR Technology that was
originally oriented to that design.

3.5.1. Design Descriptions

 

Modular steel-vessel HTR development began in The Federal Republic
of Germany (FRG) in the late 1970s. Concepts have been developed by In-
teratom, a subsidiary of Kraftwerk Union (KWU) and by Hochtemperatur
Reaktorbau (HRB).*!=*2 Kernforschungsanlage (KFA), the Nuclear Research
Center at Jlilich, has also been very active in the FRG program. They
have taken advantage of favorable HTR characteristics (e.g. high heat
capacity of the core and reflector, high temperature capability of the
fuel, large negative temperature coefficient of reactivity) to develop a
simpler plant to ease regulatory, construction and financing difficul-
ties, as well as minimize development requirements. Both the Interatom
and the HRB modular HTR concepts involve small modules of 200 to
250 MW(t) each. The thermal output of several modules can be combined
to obtain a larger total plant capacity. This approach obviously re-
duces the fission product inventory in any single reactor and reduces
the amount of heat which must be removed from a reactor core in the
event of an accident, thereby contributing to a high degree of safety.
Both concepts utilize pebble fuel as do the two existing German HTRs
[the Arbeitsgemeinschaft Versuchs Reaktor (AVR) and the Thorium Hoch
Temperatur Reaktor (THTR)]. The Interatom concept places the core and
steam generator in separate steel vessels in a side-by-side configura-
tion, while in the HRB concept the steam generator is located above the
3-46

core in the same vessel., For both concepts, the reactor vessel is
housed in a reinforced concrete cavity for both confinement and bio-
logical shielding. A vessel cooling system, mounted on the inside sur-
face of the cavity wall, is normally in operation cooling the concrete
and is capable of providing decay heat removal by heat radiation from
the uninsulated reactor vessel. Design parameters (such as core size
and power density) for these modules were judiciously combined with
generic HTR features so that in extreme accidents public safety is pro-
vided without the operation of active heat removal equipment. Engi-
neered systems are employed, but their role is primarily for investment
protection. KWU/ Interatom is actively proposing their plant design for
near-term commercial generation of electricity and for cogeneration of
electricity and process heat. HRB proposes their concept for small
electricity users and for process steam application. For the longer
term, Interatom and HRB are developing their concepts for advanced pro-
cess heat purposes such as the production of syngas through steam re-
forming of methane or by steam gasification of coal., For larger plants
HRB offers the HTR 500 [500 MW(e)].

An informal but broad survey of US utilities by Gas Cooled Reactor
Associates identified a preference for plant sizes in the 200-700 MW(e)
range for capacity additions beginning in the mid- to late 1990s.“3
Other more general studies also have indicated interest in smaller
plants. In response to these factors, the U.S. HTR Program was re-—
aligned in May 1984 to evaluate the potential for small reactor concepts
with emphases on plant investment protection and safety. 1In particular,
the plant design should be such that there would be no need for emer-
gency sheltering or evacuation of the public as a consequence of licens-
ing-basis events.

Four concepts which resulted from a preliminary screening process
were:

1170-MW(t) HIGR Cylindrical Prismatic Core Concept;
(Ref. 44)

1260-MW(t) HTGR Annular Core Prismatic Concept;
(Ref. 45)

250-MW(t) Pebble Bed Reactor Vertical-In-Line Steel
Vessel Concept (4 units of 250 MW(t) each);
(Ref. 46)

250-MW(t) Pebble Bed Reactor Side-by-Side Vessel
Concept (4 units of 250 MW(t) each).
(Ref. 35)

A Concept Evaluation Plan3* specified criteria (generally consis-
tent with NPOVS criteria) against which these plant concepts were evalu-
ated. As a result, the modular HTR in a side-by-side configuration was
selected in early 1985 as a preferred concept. Initially, emphasis was
placed on the pebble bed core concept; however, a subsequent evaluation
3-47

between pebble and prism fuels led to the selection of a prismatic core
in September 1985. The prismatic core obtains higher capacity with
power levels up to 350 MW(t) by employing an annular core design. The
higher power level reduces the plant capital cost per kW(e) for the
prismatic fueled core relative to that for the 250 MW(t) cylindrical-
core pebble bed reactor. In addition, the problem of compensating for
reactivity insertions due to water ingress is reduced in the annular
core design. The current reference modular HTR plant consists of
4 x 350 MW(t) reactor units for a total capacity of approximately 560
MW(e).

Through an integrated approach, the modular HTR concept 1is being
designed to meet the goals of safe, economical power.4?7 To meet these
goals, the design must satisfy the following requirements: 3°

1. Equivalent availability factor of 807 with planned downtime of
less than or equal to 10% per year.

2. 50 yr life measured from issuance of construction permit.

3. Have at least 107 economic advantage over the best coal-fueled
alternative source of electricity.

4, Capable of start of operation in mid 1990s.

5. Separate the nuclear and non-nuclear portions of plant to mini-
mize the number of components and systems which must be pro-
cured, installed, operated, and inspected to nuclear standards.

6. Satisfy investment protection goals:
a) less than 10% unscheduled unavailability
b) provide protection against long outages
c) limit the cost of decontamination and decommissioning
d) freguency of events leading to plant loss to be less than
1072 per plant year
7 Satisfy HTGR safety goals:

a) doses not to exceed EPA Protective Action Guidelines for
public evacuation down to an accident frequency of
5 x 10~7 per plant year

b) meet NRC interim safety goals

The reference modular HTR is shown in Figure 3.7. A plant would
consist of four 350 MW(t) reactor modules generating steam for two
nominal 300 MW(e) turbine generators to produce a net plant output of
558 MW(e) (other design alternatives using 1 x 560 kW(e) and 4 x 140
MW(e) turbines are also being examined in the DOE program to determine
the best approach).%8
3-48

ORNL-DWG 86-7363

     
 

CONTROL ROD
DRIVE ASSEMBLY

REACTOR VESSEL

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

ANNULAR
PRISMATIC CORE ——
I~
MAIN CIRCULATOR
4 1 MOTOR
N i N
: ¥ 1
‘ a1 ifl.—/— MAIN STEAM
{ — 3 C OUTLET
L } . — STEAM GENERATOR
SHUTDOWN COOLING VESSEL
HEAT EXCHANGER
UM | | —STEAM GENERATOR
SHUTDOWN COOLING 1 }
CIRCULATOR/MOTOR
v S

 

 

FEEDWATER INLET

Fig. 3.7. 350 MW(t) annular prismatic HTGR: Primary coolant flow
path during normal operation.
3-49

Fach reactor module is housed in a reinforced concrete enclosure
(silo) which is fully embedded in the earth. The nuclear island con-
gists of four enclosures and adjacent structures which house fuel
handling, helium purification, storage, and transfer systems, the rad-
waste system, nuclear island cooling water systems, and other essential
reactor service systems. A common control room is used to operate all
four reactors and the turbine plant.

Each 350 MW(t) unit consists of separate reactor and steam gener-
ator vessels connected by a horizontal coaxial crossduct. The core,
graphite reflector, core support structure, and restraining devices are
installed in the reactor vessel. The current core concept uses pris-—
matic fuel elements most of which will be geometrically identical to the
Fort St. Vrain standard (non-control) elements. The elements contain
vertical through-holes for coolant flow and blind holes for fuel rods.
The core consists of fuel elements in an annulus between an inner and
outer region of hexagonal reflector elements. A number of the elements
to be placed adjacent to the inner reflector contain an off-center hole
to accommodate the insertion of reserve shutdown materials. Although
the internal configuration of these elements differ from those used at
Fort St. Vrain, the external geometry is the same. A number of the
internal and external reflector elements which bound the core contain
off-center holes for control rod insertion. The hexagonal fuel and
reflector elements are designed for periodic replacement via the control
rod penetrations in the vessel top head. The outermost radial reflector
elements are irregular in shape so as to interface with the hexagonally
stepped outer boundary of replaceable reflector elements and the lateral
core support structure. Gravity-assisted control rod drive mechanisms
are positioned above the radial reflector to operate control rods in the
channels in the inner and outer reflector.

The active core consists of 66 10-block high columns of fueled ele-
ments. This makes the annular core configuration three elements wide
and gives an average core power density of 5.91 W/cmd3. The fuel ele-
ments contain 1.27 em (0.50 in.) diameter by 6.35 cm (2.50 in.) long
fuel rods consisting of coated UCO and THO2 particles of low enriched
uranium (LEU) fuel (U-235 < 20%) bonded in a graphite matrix. Refueling
is accomplished with the reactor shut down and the vessel depressur-
ized. The refueling operations are predicated on a three-year fuel
residence time whereby half the fuel elements are replaced at the in-
tervals of 18 months. The new fuel is placed into alternate columns
adjacent to the half-burned fuel. During refueling, all the fuel
elements in the core are moved within the vessel in 60 deg sectors at
a time; fresh and spent fuel pass through the top head refueling
penetrations which are located over the inner-reflector-to-core inter-
face. [Each sector is rebuilt with half new and half-burned fuel. At
discharge, the spent fuel ©burnup of the equilibrium cycle is
82,460 MWD/tonne. During each refueling, one-sixth of the reflector
elements adjacent to the active core is replaced which corresponds to a
nine year residence time. An alternate cycle has also been evaluated
whereby the entire core is fueled as a batch, with a lifetime of about
3-50

2.7 yr. This cycle is stated to have nearly as favorable costs and to
require less frequent shutdown for refueling.

Replacement of fuel and reflector elements is performed with the
fuel handling machine (FHM) which is placed over the inner penetration
corresponding to the sector to be removed. The FHM elevates the spent
elements into a fuel transfer cask., The fuel transfer cask, loaded to
its maximum with five elements, is used to place the elements in a fuel
storage well. Here the elements are dry-cooled before shipment off-
site. The reactor plant cooling water system is used to remove heat
from the well.

Helium flows downward through the core coolant channels to an out-
let plenum and then through the central duct of the cross duct to the
top of the steam generator. It then flows downward across the once
through helical coil steam generator with uphill boiling in the steam
generator tubes. Cool helium is drawn from the bottom of the steam gen-
erator and flows through an annulus surrounding the steam generator
outer shroud to the circulator located on top of the vessel. The circu-
lator discharges helium to a plenum from which helium flows through the
outer annulus of the cross duct to the reactor vessel. It then flows
upward through channels in the outer graphite reflector to a plenum
above the top of the core.

The reactor internal structures consist of graphite and metallic
components. The major graphite components are the outer permanent re-
flector, bottom reflector, core support posts, and top reflector. The
major metallic components are the core support plate, core barrel
lateral support structure, and the hot duct portion of the concentric
cross duct. The reactor internals are designed for the full operating
life, but are also designed to be inspectable, removable, and replace-
able, if necessary.

The main circulator, a variable speed, motor-driven single stage
centrifugal compressor using gas/magnetic bearings, is mounted verti-
cally on top of the steam generator vessel.

Design parameters are summarized in Table 3.2. The basic approach
has been to judiciously select design parameters and engineered systems
so that they combine with inherent HTR features to yield a high degree
of passive safety, and to provide investment protection as discussed in
the following paragraphs.

Two independent, diverse reactivity control/reactor shutdown sys-
tems are provided. The primary system utilizes control rods located in
the inner and outer replaceable reflector. The second system, the re-
serve shutdown system (RSS), consists of boronated graphite pellets in
storage hoppers which can be discharged into channels in the innermost
row of fuel columns. Reactivity control requirements for basic opera-
tions, including cold shutdown, are adequately covered by the reflector
rod systems alone, with at-power operations possible without insertion
3-51

Table 3.2, Summary of major design features of modular HTR
(side-by—-side configuration)

 

Power per module, MW(t)
Core power density, kW/1
Core inlet temperature, °C
Core outlet temperature, °C
Helium flow rate, kg/sec
Helium flow direction
Helium pressure, MPa (psia)
Active core diameter, m
Active core height, m

Fuel element

Fuel
Equilibrium reload, kg:U/Th
Average discharge burnup, MWd/kg

Radial reflector thickness, m

Reactor vessel material

Reactor vessel, 0D, m
Reactor vessel thickness, cm

Reactor vessel height, m

Steam condition,

pressure MPa (psia)
temperature, °C

Net thermal efficiency, %

350

5.91

258

687

156.6

downward

6.38 (925)

1.65 inner, 3.5 outer
7.8

prismatic hex—block, 20.78 cm sides
x 79.3 cm height

LEU/Th
965/881
82.5

1.0

Low alloy steel-Mn-Mo, Sa 533 GrB
Class 1

7.44
13.3
21.95

17.3 (2515)
541

39.6

 
3-52

of the inner-reflector rods. Cold shutdown with maximum positive reac-
tivity due to water ingress requires the combined insertion of the
reflector rods and the RSS. During a conduction cooldown event, the
inner control rods could be damaged because of high temperatures. To
avoid damage, although it does not affect safety, a control rod opera-
tional strategy has been adopted where the inner rods are normally used
for startup to 25% power and for normal cold shutdown.

Steam generator tube leaks are detected by a moisture monitor
located at the circulator outlet. If excessive moisture is detected,
the steam generator is isolated and dumped and the main circulator is
stopped.

A shutdown cooling system (SCS) is provided to achieve and maintain
the reactor thermal conditions required for maintainence in the event of
failure of the main heat transport system (HTS) and to help meet the
overall plant availability goal. The SCS is located in the bottom of
the reactor vessel and consists of a heat exchanger and a circulator
with a submerged motor.

The reactor cavity is provided with a natural draft air cooling
system (RCCS), Fig. 3.8. 1t consists of cooling panels mounted on the
cavity wall through which air circulates by natural convection. The de-
sign has no valves or active components. The surface of the panels
serves as a barrier separating the outside atmosphere from the reactor
cavity atmosphere. The system uses four separate inlet/outlet struc-
tures to minimize the possibility of flow blockage. 1In addition, the
four loops are interconnected by inlet/outlet plenums in the cooling
panels. This provides a heat sink sufficient for decay heat removal in
the event the main heat transport system (steam generator and main
helium circulator) and the shutdown cooling system are not available.
Heat transport from the reactor core is by natural processes of conduc-
tion and radiation (and convection if the primary system is pressurized)
through the core to the vessel wall and by radiation and convection to
the cooling panels.

The reactor utilizes a confinement equipped with dampers which open
on excessive pressure loads resulting from feedwater, main steam, or re-
actor coolant line ruptures. Program studies indicate that the fission
product releases from the core are small enough that reliance need not
be placed on conventional pressure~tight containment or a confinement
with a filter system to meet the defined safety criteria.

For decay heat removal, under pressurized or depressurized condi-
tions, the main cooling loop (consisting of the main circulator and the
steam generator) is the first option. If either the main circulator or
the steam generator is not operational, then forced circulation using
the shutdown cooling system is the next option for either pressurized or
depressurized conditions. The next option is to remove decay heat
through the vessel wall by radiation to the RCCS. This system is
designed to limit the fuel temperatures to 1200°C under pressurized
conditions (when there can be a significant redistribution of heat
3-53

ORNL—-DWG 864054 ETD

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4.
..l
Feoanr

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=

. 2%,

 

 

 

 

 

T avte "

v
1.t

 

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LI
-

 

 

 

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{2ee.

 

 

FEATURES :

1. COOLS THROUGH PANEL WALL
2 OPERATES UNDER ALL MODES OF REACTOR OPERATION.
3. SIMPLE TO OPERATE, VIRTUALLY MAINTENANCE FREE.

Fig. 3.8. Reactor cavity cooling system.
3-54

within the core by natural convection) and to 1600°C under depressurized
conditions.

3.5.2 Claims, Advantages, and Disadvantages Evaluated Against Criteria,

Essential and Desirable Characteristics

 

The claims and reported advantages of the modular HTR are discussed
briefly as follows in the order of the criteria first and essential and
desirable characteristics second. A more detailed evaluation of modular
HTR claims has been included in Appendix G.

1.

Public Risk: While the calculated risk to the public has not
been quantified for the modular HTR, there are important fea-
tures which provide the design with a high degree of passive
safety, and thereby also provide confidence that the calculated
risk to the public due to accidents will be equal to or less
than the calculated risk associated with the best modern
LWRs. These features are:

® The capability for afterheat removal through the vessel
wall by natural heat transport mechanisms (convection,
conduction, and radiation). This capability has been
demonstrated partially on the smaller experimental AVR
which shares features of the modular HTR. Future
confirmatory experimentation for more generally applicable
data is being considered and may be possible at the AVR
subject to the approval of the German authorities.

® Very good retention of fission products within the fuel to
high temperatures. This feature has been demonstrated by
U.S. and German coated particle fuel systems exXperience
and in fuel test programs.

° No need for a fast acting shutdown system for core heatup
events, which again has been demonstrated on other HTRs.

Other potentially severe accidents, such as major water
and/or air ingress events, have been argued to be of such low
consequence or low probability by virtue of system design that
these types of accidents pose no significant public risk.
However, the NRC may require that these accidents be factored
into the cost benefit analysis of the use of confinement versus
containment.

Investment Risk: The probability of loss of investment for the
modular HTR is claimed to be less than 107° per vyear. This
claim requires independent review, but many of the features
which preclude or reduce the effect of incidents on public
safety can be argued as being favorable to investment protec-
tion.
3-55

Economic Competitiveness: Until the plant design is complete
and commodity requirements determined, a firm estimate of cost
cannot be made. With regard to meeting the financial goals of
the utility, the ability to add capacity in small increments as
well as the potential for achieving short construction time
through factory fabrication should reduce the utilities' capi-
tal investment exposure and investment risk, thereby helping to
meet their financial goals. With regard to acceptable busbar
costs, factory fabrication of modules coupled with the rela-—
tively high burnup achievable with HTR fuel cycles may com-
pensate for higher fuel fabrication costs typical of HTRs and
potentially higher distributed capital cost usually associated
with smaller sized plants. The use of multiple modules may
also increase overall availability, although at lower power
levels, thereby providing flexibility in scheduling outages.
Assumptions about availability for the modular plants play an
important role in estimating overall competitiveness with both
the coal fired and the better current generation LWR plants.

Probability of Cost/Schedule Overruns: The DOE and the in-
dustrial proponents recognize the need for complete design
before initiating construction. Detailed design and associated
studies of construction needs, options and costs still remain
to be completed, so that cost and schedule factors cannot be
quantified. However, the DOE funded program has produced
indepth studies of construction needs, options, and costs for
the modular HTR so that uncertainties should be well defined.

Licensability: The modular HTR has a draft licensing plan.
The DOE and industrial proponents are actively engaged in
dialogue with NRC-NRR. Early concurrence on licensing may be
essential to meeting the 2000-2010 time frame for commercial-
ization. DOE and the industrial proponents plan to secure an
NRC final design approval (FDA) by 1996. A preliminary safety
information document (PSID) will be submitted in CY 1986.
There is also a utility effort led by the Tennessee Valley
Authority (TVA) which proposes joint funding of a single plant
to demonstrate licensability by test; however, such testing
probably could not address all safety questions particularly
those beyond design basis accidents such as a major air
ingress, acts of sabotage and selsmic events.

Demonstration of Readiness: Many features of the modular HTR
are or will have been demonstrated in the German AVR plant
before the modular HTR is offered commercially. A successful
demonstration of high powered, gas/magnetic bearing circulators
would represent a significant contribution to the demonstration
of readiness since most of the other major component tech-
nologies either are borrowed or have evolved from AVR, THTR,
and Fort St. Vrain experience. More fuel testing is already
planned to support licensing as well as normal operation
requirements.
3-56

Owner Competence: The operation of multi-module plants may
pose new concerns about interdependence and common mode inter-
actions of systems. Such concerns may influence NRC mandates
on acceptable control configurations which may in turn be more
costly or manpower intensive than currently envisioned; how-
ever, the overall technology of the modular HTR appears to be
as readily assimilable as IWR technology. The lessons from the
Fort St. Vrain HTGR also appear to be clear to a potential
owner/operator of an HTR. Some of these lessons are: (1) keep
moisture out of the primary system and any other part of the
plant where it can cause corrosion [which should be helped in
the modular design by incorporating hardware features based on
lessons learned to date], (2) maintain excellent secondary
coolant chemistry, (3) wmaintain an intensive and extensive
surveillance program of (1) and (2) above, (4) ensure quality
and confirmatory testing of both original and replacement
materials and equipment, and (5) maintain a clean physical
plant. The large thermal margins inherent in the HTR fuel sys-
tems and the low graphite corrosion rates in the presence of
numerous moisture ingress events at Fort St. Vrain could 1lull
plant designers and operators into a failure to recognize the
significance of operating problems. Some of the observed -oper—
ational problems at Fort St. Vrain have included the effect of
moisture on leaching and distribution of other corrosive mate-
rials (e.g., chlorides), the apparent inability to detect ab-
normal control configurations and reactivity anomalies quickly
(e.g., confirming subcriticality by excore detectors and
detecting dropped control material) and the possible inter-
dependence of redundant emergency ac power sSystems. These
types of situations should not be repeated with the modular
HTR.

Essential Characteristics: Many of the essential character-
istics are integral requirements for meeting one or more of the
criteria and as such are discussed more fully above. However,
in general, the modular HTR has promise of achieving many of
these characteristics as outlined, in some cases repetitively,
below:

a. High availability due to use of small-sized turbines and
modularity which allows higher availability at reduced
power.

be Maximum use of shop fabrication of reactor systems

Ce A high degree of passive safety

d. Potentially no need for developing or demonstrating a plan
for evacuation of the public beyond the site boundary

e. Potential for demonstrating features important for passive
safety
The

3-57

f. Low thermal discharge (due to high thermal efficiency)

g Low radiocactive effluent as demonstrated by Peach Bottom
1, Dragon, Fort St. Vrain and AVR experience

h. Low investment risk to the utility resulting from adding
capacity in small increments and from what is intended to
be a simpler approach to meeting safety requirements and

licensing.
Desirable Characteristics: Several of these characteristics
are addressed in regard to criteria. The advantageous omnes are
listed again as follows. The modular HTR appears to have

modest RD&D requirements, relative ease in siting based on pro-
jected low source terms for both normal operation (worker expo-
sure and effluents) and accident conditions, good fuel
utilization (high burnup), high thermal efficiency, high ver-
satility in application because of the production of high
coolant temperatures, and a low visual profile through full
embedment. Full embedment and passive safety should also con-
tribute to a high degree of sabotage resistance. The claim is
made that the use of low enriched uranium increases resistance
to proliferation and diversion and that appears to be the case
for the fresh fuel supply. Also, the HTR spent fuel appears to
have a high resistance to diversion and proliferation tech-
nologies.

potential disadvantages are discussed as follows:

Public Risk: As alluded to under the discussion of advantages,
the resolution of concerns for severe accidents will preferably
be handled by probabilistic risk analyses to demonstrate a low
contribution to the overall risk to the public. If require-
ments such as the use of inerted containment are imposed, over-
all costs will increase.

Investment Risk: Independent assessments are needed.

Economic Competitiveness: The possibility of higher costs for
fuel fabrication and plant capital investment are a concern; as
is the availability which will be achieved. Independent evalu-
ations appear prudent to perform.

Probability of Cost/Schedule Overruns;: Since this 1is a new
concept, it is particularly important that the design is com—
pleted before construction begins. The current approach is
based on defining "top-down" requirements from which design
data needs and RD&D will proceed using functional analysis.
Construction plans and schedules must be coordinated carefully
with the availability of design and safety related data.

Licensability: As mentioned above, the analysis requirements
and expected design needs in response to "beyond design basis
3-58

accidents” must be settled, preferably, as early as possible in
the design process. 1In the post-TMI licensing environment, the
modular HTR could still face defense-in-depth requirements such
as containment, emergency ac power sources and safety grade
components in the balance of plant. These can have a severe
effect on increasing plant costs if imposed. Seismic consider-
ations with regard to the reactor core and the side-by-side
connecting pipe must also be addressed for licensing.

6. Demonstration of Readiness: Other than answering questions
about needing high availability for overall competitive eco-
nomics, the modular HTR would appear to have a lesser require-
ment for a demonstration plant because of AVR and THTR experi-
ence and the ability to incorporate lessons learned at Fort St.
Vrain.

7. Owner Competence: No specific disadvantage identified, how-
ever, as indicated under advantages, a potential owner/operator
should be thoroughly familiar with details of the engineering
and licensing experience at St. Vrain. The lessons learned are
positive with respect to avoiding potential pitfalls.

8. Essential Characteristics: The relatively low power per module
[~350 MW(t)] does affect the capital cost as a disadvantage.
The side-by-side HTR module has also been questioned because of
potentially adverse seismic response at the connecting pipe be-
tween the reactor and steam generator vessels. Both of these
features may be improved through design enhancement and innova-
tion. The power of the module may be increased if higher fuel
temperatures (>1600°C) become acceptable by further fuels test-—
ing and verification. The connecting pipe will require
thorough and extensive analysis to show that it can withstand
the potential consequences from seismic events.

9. Desirable Characteristics: The use of LEU/Th fuel leads to

lower fuel conversion ratios relative to the use of highly en-
riched fuels.

3.5.3 Modular HTR Research and Development Needs Evaluated

 

Within the DOE HTR Program, development of a modular HTR Technology
Development Plan using the Integrated Approach is under way, but results
are not sufficiently complete for incorporation into NPOVS. However,
ORNL has prepared a document“? which was presented to the Subcommittee
on Energy Research and Production of the U.S. House of Representatives
and which discussed the key research and development areas required for
modular HTRs. This section presents R&D needs as excerpted from this
document with modifications reflecting additional information obtained
since that time. The key research and development (R&D) areas are con-
sidered in the following categories:
A.
B.
C.

3-59

Base Technology
Applied Technology; and
Design and Economic Studies.

Item A generally refers to basic information needed to establish
the feasibility of the reactor concept and to materials data needed for
the detailed design; item B refers to R&D needed to assure the practi-
cality of components and systems; and item C refers to the effort re-
quired to specify the entire reactor plant in sufficient detail to
permit reliable economic estimates of plant performance.

The key R&D areas which need to be addressed for the modular HTR
are shown below:

3.5.3.1

1.

8.

3.5.3.2

1.

Base Technology

 

Determination of fission product retention of the fuel coat-
ings, graphite and metal surfaces of the primary system and
confinement during and subsequent to extreme accident condi-
tionse.

Process development for fuel fabrication and irradiation test-
ing to obtain understanding of the importance of specific pro-
cessing parameters on fuel performance.

Irradiation testing and examination of fuels produced in com—
mercial-scale production equipment.

Fission product behavior during normal reactor operation as
related to lift-off and the source terms under depressurization
accidents.

Development of detailed materials properties under conditions
of creep, fatigue, corrosion, and radiation necessary for
designing and operating components.

Obtaining statistical data on graphite properties as a basis
for estimating fuel element stresses.

Critical experiment testing of LEU/Th cores, including water
ingress reactivity effects and temperature coefficients for low
enriched uranium fuel with plutonium concentrations representa-—
tive of equilibrium burnup.

Obtaining experimental data to validate codes applicable to the
passive heat removal system.

Applied Technology

 

Development, verification, and application of analytical tools
for reactor design, safety, and risk analyses, including data
bases.
3-60

2. Plant safety and risk analyses. This risk associated with
normal operation and design basis accidents needs to be inves-
tigated. Also, risk associated with postulated events beyond
design basis may need to be investigated to confirm that the
total risk from such accidents is small relative to the risk
from normal plant operation.

3. Detailed reactor physics analysis including computation of
cross sections, power distributions, temperature coefficients,
and control rod worth under normal conditions and with water
ingress. Also shielding analysis to determine fluence for the
design of reactor internal components at various locations.

4. Design and testing of refueling equipment to demonstrate that
the reference reactor concept can be refueled on the assumed
schedule,

5. Design and testing of prototypic components and systems such as
the helium circulator, core support structure, and shutdown

cooling heat exchanger.

6. Development of multi-module control system, service systems,
and heat exchange systems.

3.5.3.3 Design and Economic Studies

 

The design of the modular HTR plant must be completed in sufficient
detail to permit a firm estimate of plant costs, based on features which
limit fuel temperatures under accident conditions, facilitate shop
fabrication, and reduce balance-of-plant (BOP) costs. Also, a detailed
determination of operating and maintenance and fuel cycle costs are re-—
quired.
3.6

l.

5.

10.

3-61

REFERENCES FOR CHAPTER 3

K. Hannerz (ASEA-ATOM), Towards Intrinsically Safe Light Water Re-

 

actors, ORAU/IEA-83-2 (M)-Rev. (research memorandum), Institute for

Energy Analysis, Oak Ridge Associated Universities, 0Oak Ridge, Ten-—
nessee, July 1983.

SECURE P: Design Progress Information, ASEA-ATOM, Vasteras, Sweden

 

(April 1984). ASEA~ATOM PROPRIETARY.

D. Babala and K. Hannerz, "Pressurized Water Reactor Inherent Core
Protection by Primary System Thermohydraulics,” Nuclear Science and

 

Engineering 90(4), 400—410 (August 1985).

C. Sundqvist and T. Pederson, "PIUS: The Forgiving Reactor, Safety
and Operational Aspects,” full text of paper presented at the 1985
Annual Meeting of the American Nuclear Society (ANS), Boston,
Massachusetts, June 9—14, 1985; ASEA-ATOM, Vasteras, Sweden.

Je. De Duncan and C. D. Sawyer, "Capitalizing on BWR Simplicity at
Lower Power Ratings,” SAE Technical Paper Series 859285 reprinted
from p. 164, Proceedings of the 20th Intersociety Energy Conversion

 

Engineering Conference, Miami Beach, Florida, August 18-23, 1985,

 

General Electric Company, San Jose, California.

Lyle C. Wilcox, "U.S. Department of Energy Programs on Cost Reduc-—
tion,” presented at the Institute of Applied Energy International
Symposium on LMFBR Development, Tokyo, Japan (November 7, 1984).

Alan E. Walter and Albert B. Reynolds, Fast Breeder Reactors, Per-
gamon Press (1981).

 

Transactions of the ANS 1984 Winter Meeting 47, 13-16 (Novem-

 

ber 11-16 1984).

Consolidated Management Office for the IMFBR of the Electric Power
Research Institute, LSPB Design Descriptions, Vol. 1, CDS 400-8,
for the U.S. Department of Energy, September 1984. APPLIED
TECHNOLOGY .

 

R. A. Lindley, Large Scale Prototype Breeder Cost Effectiveness

 

Considerations, Consolidated Management Office for the IMFBR of the

Electric Power Research TInstitute, August 1984. APPLIED

TECHNOLOGY.
11,

12.

13,

14,

15.

16.

17.

18.

19.

20.

21.

22.

23,

24,

3-62

Consolidated Management Office for the LMFBR of the Electric Power
Research Institute, LSPB Overall Plant Design Specification, CDS
100-2, Rev. 7, for the U.S. Department of Energy and Electric Power
Research Institute, Washington, D.C., February 1984, APPLIED
TECHNOLOGY.

LSPB Constructibility Report, to be published by the U.S. Depart-

 

ment of Energy. APPLIED TECHNOLOGY.

Modular IMFBR Pool Plant Final Report, AI-DOE-13502, Rockwell In-

 

ternational, Atomics International, Canoga, Park, California,
September 30, 1984, APPLIED TECHNOLOGY.

"Advancing Breeder Reactor Design in the United States,”
Engineering International 30(365), 17-20, (February 1985).

Nuclear

 

Transactions of the ANS 1984 Winter Meeting 47, 299-300 (November

 

11-16, 198%4).

“SAFR discussions at ORNL, January 11, 1985," a collection of view-
graphs presented at this meeting., APPLIED TECHNOLOGY.

Large IMFBR Pool Plant, Vol. 1, Design Description, ESG-DOE-13410

 

Rockwell International, Energy Systems Group, Canoga Park, Cali-
fornia, September 1983. APPLIED TECHNOLOGY.

PRISM Semiannual Report, April-September, 1984, XL-897-840073/L3,

 

General Electric Co., Nuclear Systems Technology Operation, Sunny-
vale, California, October, 1984. APPLIED TECHNOLOGY.

Jo S¢ Armijo et al., "General Electric Strategy for Achieving a
Low—Cost Liquid Metal Reactor Plant,” Presented at the Institute of
Applied Energy International Symposium on ILMFBR Development, Tokyo,
Japan, November 7, 1984,

"Advancing Breeder Reactor Design in the United States,”
Engineering International 30(365), 17-20 (February 1985).

Nuclear

 

PRISM Design Requirements, Preliminary Rev B, 23A3071, General
Electric Co., Nuclear Systems Technology Operation, Sunnyvale,
California, October 1984, APPLIED TECHNOLOGY.

Internal Correspondence from G. F. Flanagan to Distribution, "In-
herently Safe LMRs,” December 6, 1984,

Clinch River Breeder Reactor Plant Probabilistic Risk Assessment,
prepared for the U.S. Department of Energy by Technology for Energy
Corporation, September 14, 1984,

A. Bayer and K. Koberlein, "Risk-Oriented Analysis on the German
Prototype Fast Breeder Reactor SNR-300," Nuclear Safety 25(1), 30
(January-February 1984).
25,

26.

27.

28.

29 .

30,

31.

32.

33.

34.

35.

36.

37.

3-63

Transactions of the ANS Winter Meeting 47, 333-338, (November
11-16, 1984).

 

"Looking to the Future with the Integral Fast Reactor,’
Engineering International 30(365), 20 (February 1985).

 

R. Balent and J. Yedidia, DRAFT, Large Scale Prototype Breeder Fuel
Cycle Plan, to be published by the U.S. Department of Energy.
APPLIED TECHNOLOGY

 

LSPB Research and Development Requirements, CDS 500-6, U.S. Depart-—

 

ment of Energy, Washington, D.C., September 1984, APPLIED
TECHNOLOGY.

SAFR Requirements for Base Technology Program, 149T1000002, Rock-
well International Rocketdyne Division, Canoga Park, California,
January 1985, APPLIED TECHNOLOGY.

Letter number XL-897-850016 from L. N. Salerns to Francis X#
Gavigan, dated January 11, 1985, "WBS2B(O.5- Initial PRISM R&D Re-
quirements Statements.” APPLIED TECHNOLOGY.

Letter from J. Ray to Dr. Bill Harms, dated January 31, 1983, with
the attachment, "LMFBR Safety Philosophy Issues”, Advanced Reactors
Subcommi ttee, draft, January 27, 1983.

Letter number T-85-053 from J. D. Mangus to D. C. Gibbs, dated July
23, 1985, with attachment, "Long Life Liquid Metal Core Concept.”

Utility/Users Design Requirements for Small High Temperature Gas-
Cooled Reactors, GCRA 84-011, Gas—Cooled Reactor Associates, San

 

Diego, California, November 1984. APPLIED TECHNOLOGY.

HTGR Program Concept Evaluation Plan for Small HTGRs, GCRA 84-009,
Gas—-Cooled Reactor Associates, San Diego, Califormia, October 31,
1984+ APPLIED TECHNOLOGY.

Preliminary Concept Evaluation Report, 4 x 250 MW(t) HTGR Plant
Side~by—-Side Steel Vessel Concept, HTGR-85-005, issued by Bechtel
Group, Inc., et al., for Gas—Cooled Reactor Associates, San Diego,
California, February 1985. APPLIED TECHNOLOGY.

FY 1985 HTGR Summary Level Program Plan, HP-20202-85, Gas-Cooled
Reactor Associates, San Diego, California, October 1984, APPLIED
TECHNOLOGY.

 

Licensing Plan for the Standard HTGR (Draft), GCRA 85-001, Bechtel
Group, Inc., et al., January 1985.
38.

39.

40.

41,

42,

43.

44,

45.

46,

47.

48.

3-64

Preliminary Concept Description Report, 4 x 350 MW(t) HTGR Plant
Side-by—~-Side Steel Vessel Prismatic Core Concept, HTGR-85-142,
issued by Bechtel Group Inc. for Gas—Cooled Reactor Associates, San
Diego, California, October 1985. APPLIED TECHNOLOGY.

P. R. Kasten et al., Assessment of the Thorium Fuel Cycle in Power
Reactors, ORNL-TM-5565, Oak Ridge National Laboratory, Oak Ridge,
Tennessee, January, 1977.

P. R. Kasten, "Statement on an Inherently Safe High-Temperature
Gas—Cooled Reactor Program,” presented to Subcommittee on Energy
Research and Production, U.S. House of Representatives, M. Lloyd,
Chairman, February 7, 1984.

He Reutler and G. Lohnert, "The Modular High-Temperature Reactor,”
Nuclear Technology, 62(1), 22-30, (July 1983).

 

HTR 100-MW(e) Konzeption; Technik, Termine, Kosten, Hochtemperatur
Reaktorbau, Hochtempatur Reaktorbau, Mannheim, Federal Republic of
Germany.

 

Summary Report on the Utility Industry Questionnaire, GCRA.84-001,
Gas Cooled Reactor Associates, San Diego, California, February
1984.

 

Preliminary Concept Evaluation Report, 1170 MW(t) HTGR Plant PCRV

 

Concept, HTGR-85-004, issued by Stone & Webster Engineering Corpor-
ation for Gas—Cooled Reactor Associates, San Diego, California,
February 1985. APPLIED TECHNOLOGY.

Preliminary Concept Evaluation Report, 1260 MW(t) HTGR Plant PCRV

 

Concept, HTGR-85-003, issued by Stone & Webster Engineering Corpor-
ation for Gas-Cooled Reactor Associates, San Diego, California,
February 1985. APPLIED TECHNOLOGY.

Preliminary Concept Evaluation Report, 4 x 250 MW(t) HTGR Plant In-
Line Steel Vessel Concept, HTGR-85-006, 1issued by Bechtel Group,
Inc., et al., for Gas—-Cooled Reactor Associates, San Diego, Cali-
fornia, February 1985. APPLIED TECHNOLOGY.

 

An Integrated Approach to Economical, Reliable, Safe Nuclear Power
Production, ALO-1-11, Combustion engineering, Inc., Windsor,
Connecticut, June 1982.

Turbine Selection Trade Study 4 x 250 MW(t) HTR Plant SBS/SV Con-

 

cept, HTGR-85-075, Stone and Webster Engineering Corporation for
Gas-Cooled Reactor Associates, San Diego, California, July 1985.
APPLIED TECHNOLOGY.
4. ACKNOWLEDGMENTS

The form and scope of this study necessitated the involvement of
many individuals and organizations. 1In fact, the numbers are so great
and the involvement so often indirect that complete individual recogni-
tion is next to impossible. However, the cooperation was extensive and
effective; those listed as authors recognize and greatly appreciate this
assistance. The institutions and individuals who contributed through
interview and/or written reports and, in some cases, through work
specific to the study are as follows:

Reactor Vendors

 

ASEA-ATOM

Babcock and Wilcox

Combustion Engineering

GA Technologies

General Electric Company

Rockwell International

Westinghouse—~Advanced Energy Systems Division

Architect-Engineers

 

Bechtel

Sargent and Lundy

Stone and Webster

United Engineers and Constructors

Utility Companies and Associations

 

Baltimore Gas and Electric

Central Electricity Generating Board, UK

Carolina Power and Light

Duke Power Company

Flectric Power Research Institute

Electric Power Research Institute Counsolidated Management Office
Gas-Cooled Reactor Assoclates

Houston Power and Lighting

Southern California Edison

Wisconsin Electric Power Company

Laboratories, Institutions, and Universities

 

Argonne National Laboratory

Atomic Industrial Forum

Institute for Euergy Analysis
International Atomic Energy Agency
Los Alamos National Laboratory
Massachusetts Institute of Technology
Nuclear Energy Agency

Office of Technology Assessment

The University of Tennessee

U.S. Nuclear Regulatory Commission

4-1
4=2

Individuals at the three cooperating institutions (ORNL, TVA, and
the University of Tennessee) who provided assistance include the follow-
ing:

Qak Ridge National Laboratory

 

S. J. Ball R. M. Harrington
T. E. Cole W. O. Harms

R. M. Davis J. E. Kibbe

J. C. Ebersole (consultant) 0. H. Klepper

J. R. Engel A. E. Levin

G. F. Flanagan G. Samuels

L. C. Fuller J. We Sims

S. R. Greene

Tennessee Valley Authority

D. T. Bradshaw J. G. Stewart
D. L. Lambert R. E. Taylor
H. G. O0'Brien S. Vigander

J. E. Simmons

The University of Tennessee

 

H. L. Dodds, Jr.

NPOVS Advisory Committee Members

 

S. Burstein, Vice-Chairman of the Board, Wisconsin Electric Power
Company

G. F. Dilworth, Director of Engineering and Technical Services
(DETS), TVA

T. S. Elleman, Vice President of Corporate Nuclear Safety and
Research, Carolina Power and Light Company

P. R. Kasten (Secretary), Technical Director, Gas-Cooled Reactor
Programs, ORNL

L. M. Muntzing, Doub and Muntzing, Washington, D.C.

D. R. Patterson, Assistant to Manager, Office of Engineering Design
and Construction, TVA

W. T. Snyder, Dean, College of Engineering, The University of
Tennessee

J. Taylor, Vice President and Director, Nuclear Power Division,
EPRI

N. E. Todreas, Department of Nuclear Engineering, Massachusetts
Institute of Technology
4-3

NPOVS Advisory Committee Members (continued)

 

J. Taylor, Vice President and Director, MNuclear Power Division,
EPRI

N. E. Todreas, Department of Nuclear Engineering, Massachusetts
Institute of Technology
APPENDIX A

BASIC OUTLINE FOR NUCLEAR POWER OPTIONS VIABILITY STUDY

VOLUME 1

VOLUME II

VOLUME TI1I

VOLUME IV

FINAL REPORT

EXECUTIVE SUMMARY

REACTOR CONCEPTS, DESCRIPTIONS, AND ASSESSMENTS
(see page v)

NUCLEAR DISCIPLINE TOPICS

ABSTRACT

1. INTRODUCTION

2. CONSTRUCTION

3. ECONOMICS

4. REGULATION

5. SAFETY AND ECONOMIC RISK

6. NUCLEAR WASTE TRANSPORTATION AND DISPOSAL
7. MARKET ACCEPTANCE

8. ACKNOWLEDGMENTS

APPENDIX A. INTERVIEW FORMAT FOR THE ISSUE DEFINITION
RESEARCH AND OUTLINE OF ISSUES USED FOR THE
CASE STUDY INTERVIEWS

APPENDIX B. TABLES ON THE SAMPLE USED FOR THE ISSUE
IDENTIFICATION RESEARCH

BIBLIOGRAPHY

ABSTRACT

1. INTRODUCTION

2. ORGANIZATION AND RETRIEVAL

3. KEYWORD LIST

4 KEYWORD INDEX

5. NUCLEAR OPTIONS CITATIONS

6. LIGHT WATER REACTORS CITATIONS
7. LIQUID METAL REACTORS CITATIONS
8. HIGH TEMPERATURE REACTORS CITATIONS
9. ACKNOWLEDGMENTS
10. REFERENCES
APPENDIX B

THE OUTLOOK FOR ELECTRICITY SUPPLY AND DEMAND*

The principal determinants of future electricity demand will prob-
ably be the utilities and their regulators. During the past ten years,
utilities have been evolving from a supply industry concerned only with
meeting electricity requirements to a service-oriented industry con-
cerned not only with the supply of electricity but also with controlling
and shaping its use through conservation and load management. Future
electricity use will depend on how far this evolution proceeds.

The approach taken to estimate future energy use involves an anal-
ysis and/or estimate of the trend of factors that determine energy use,
such as population, persons per household, gross national product (GNP),
shifts in the industrial product mix, conservation, etc. The projec-
tions made here do not represent anything even approaching the tech-
nology limits of energy conservation nor do they come close to the eco-
nomic 1limit of conservation as projected by "least cost energy
strategies.” They do depend on continued efficiency improvements and,
to some extent, on a coutinuation of utilities' aversion to investment
in new capacity, which has resulted in conservation and load management
programs to limit demand growth. They probably represent a narrow band
in the upper part of a rather wide range that could be expected.

Table B.}l summarizes the estimates of this study for growth rates
of electricity and nonelectrical energy requirements to the year 2000
for the residential, commercial and industrial sectors. The total
growth rate for electricity is estimated to range between 1.8 and 2.3%/
year and for nonelectrical energy between 0.1 and 0.5%/year. These
rates result in a growth of primary energy requirements of 0.9 to 1.4%/
year, which is equivalent to using between 67.3 and 73.9 quads (exclud-
ing tramsportation) in the year 2000. The transportation sector is not
analyzed in this study since this sector does not use a significant
amount of electricity and, barring a breakthrough in battery technology
is expected to use very little electricity for the remainder of the cen-
tury.

The residential sector projections are based on the following
assumptions: (1) a population growth rate (as projected by the Bureau
of the Census) of 0.85%/year between 1980 and 2000); (2) a household
growth rate of 1.4%/year, which would continue the trend of households
growing at a rate about 607 greater than the population; (3) a continu-
ation, at a modest rate, of the trend to less energy use per household;
and (4) a continuation of the trend to electric space heating.

 

*Taken from G. Samuels, The Outlook for Electricity Supply and
Demand, ORNL/TM-9469, 0Oak Ridge National TLaboratory, Oak Ridge,
Tennessee, April 1985,

 
Table B.l.

residential,

Projected energy use for the
comnercial, and industrial
sectors in the year 2000

 

 

1980-2000 End use Primary
Sector annual growth energy energy use
(%/year) (1013 Btu/year) (1015 Btu/year)
Residential
Electricity 1.50 to 2.00 3.30 to 3.64 11.29 to 12.45
Nonelectricity -1.50 to -1.00 5.09 to 5.63 5,09 to 5.63
Total primary 16,38 to 18.08
Commercial
Electricity 2.00 to 2,50 2.83 to 3.12 9.70 to 10.69
Nonelectricity 0 4,09 4,09
Total primary 13.79 to 14,78
Industrial
Electricity 2,00 to 2.50 4,13 to 4.56 14.15 to 15.60
Nonelectricity 0.50 to 1.00 22.99 to 25.39 22,99 to 25.39
Total primary 37.14 to 40.99
U.S. total
Electricity 1.83 to 2.33 10,26 to 11.32 35.14 to 38.74
Nonelectricity 0,06 to 0.50 32.17 to 35.11 32.17 to 35.11
Total primary 0.90 to 1.37 67.31 to 73.85

 
B-3

The commercial sector projections are predicated on a substantial
decline in the growth rate of both sectoral employment and floor
space—to an annual rate of 1.5%Z. Electricity use per employee or per
unit of floor space was assumed to increase at a rate 0.5 to 1.07%
greater than employment or floor space.

The industrial sector projections are based on a detailed analysis
of the manufacturing industries between 1975 and 1980, which examined
changes in the energy intensity and output of these industries at the
four—-digit Standard Industrial Classification level. Electricity use
for these industries is projected to grow at a rate equal to about 80%
of the gross national product growth rate, which is expected to be in
the range of 2.5 to 3.0%Z for the remainder of the century.

Although these estimates are small compared to most projections of
several years ago, they are in the range of recent projections and close
to current "conventional wisdom.” An examination of past energy use
suggests that the rapid growth between 1950 and 1970 was self limiting
and that the o0il price shocks of the 1970s were a catalyst that ended
this rapid growth. The technologies that led to this growth were avail-
able by 1930. However, the Depression and World War II delayed their
growth, which resulted in their impact being compressed into a shorter
time span and the rapid growth of the 1950's and 1960's.

The utilities' projections of future demand and their plans for
future generating capacity have declined steadily over the past ten
years., Projections for peak demand and electrical energy requirements
in 1992 represent a 2.25 and 2.61%/year growth from actual 1980
values. Their projections indicate that average reserve margins for the
contiguous United States should be adequate through 1992. Reserve mar-
gins are projected to decline slowly from 417 in 1982 to 30% in 1992.
Furthermore, based on utility projections, each of the nine regional re-
liability councils will have reserve margins of at least 207 in 1992.
However, the adequacy of both regional and U.S. electricity supply
depends primarily on the validity of the drastically reduced projections
of future demand growth and to a lesser extent on the utilities' ability
to provide the planned generating capacity. For example, if utilities
were to complete only those units now under construction and if demand
grows as projected, 1992 reserve margins would be 22 to 23%. However,
if demand were to reach that projected in 1980 (a 47 annual growth
rate), completion of all currently planned capacity by 1992 would pro-
vide only a 6% margin—far too small to maintain service during peak
demand periods.,

The sensitivity of reserve margins to the demand growth rate, com-
bined with a long lead time required to add economical capability, has
led to concerns about the adequacy of future electricity supply. At the
same time consumer resistance to higher electricity prices and the re-
sulting pressure on Public Utility Commissions has seriously affected
the utilities' ability to finance the capacity now being built. Adding
more capacity as insurance for an unexpected increase in demand would be
difficult to sell to either consumers or utilities at this time.
B-4

Relatively low—cost approaches exist for lessening the probability
of future electricity shortages. One approach would be to allow advance
siting and permitting and then "banking” of sites so that the lead time
would be reduced to that required for constructiom—about half of the
current 8- to l2-year lead time. The time for which the construction
permit remains valid would have to be increased.

A second approach would follow a path now being adopted by a few
utilities. This approach would treat conservation and load management
as supply options. Utilities would, with the approval of regulators,
channel capital into the most economical option to meet future service
requirements whether this option be increased capacity or reduced de-
mand. Treating demand-reducing options as a supply would permit
“"capacity” addition to more closely match increases in demand. Further-
more, this option would provide results in less time than that required
for adding large central stations. This shorter lead time would also
alleviate the debate over including construction work in progress in the
rate base.

REFERENCES [Used in G. Samuels, The Outlook for Energy Supply and

 

Demand, ORNL/TM-9469, 0Oak Ridge Natioumal Laboratory, Oak Ridge,
Tennessee 37831 (April 1985).]

A. P. Sanghvi, "Least Cost Energy Strategies for Power System Expan-
sion,” Energy Policy 12(1),75-92 (March, 1984).

R. H. Williams, G. S. Dutt, and H. S. Geller, "Future Energy Savings in
U.S. Housing,” Annual Review of Energy 8, 269-332 (1983).

 

Survey of Utility Load Management and Energy Conservation Projects,
EPRI/EM-1606, FElectric Power Research Institute, Palo Alto, Calif.,
November 1980.

Conference Proceedings Utilities and Energy Efficiency; New Opportun-
ities and Risks, October 23-24, 1980, CONF-8010146, , Port Chester, N Y.

 

State Energy Data Report, 1960 through 1980, DOE/EIA-0214(80), U.S.
Department of Energy, Washington, DC, July 1982.

Statistical Abstract of the United States, 1982-83, U.S. Department of
Commerce, Bureau of the Census, Washington, DC.

 

Residential Energy Consumption Survey: Consumption and Expenditures
April 1980 through March 1981, DOE/EIA-0321/1, U.S. Department of
Energy, Washington, DC, September 1982.

 

Residential Energy Consumption Survey: Housing Characteristics 1980,
DOE/EIA-0314, U.S. Department of Energy, Washington, DC, July 1982.
B-5
Residential Energy Conservation, Volume I, OTA-E-~02, U.S. Congress,

Office of Technology Assessment, Washington, DC, July 1979.

1982 Annual Energy Outlook with Projections to 1990, DOE/EIA-0383(92),
U.S. Department of Energy, Washington, DC, April 1983.

The Future of Electric Power in America: Economic Supply for Economic
Growth, DOE/PE-0045, Department of Energy, Washington, DC, June 1983.

 

J. F. Gustaffero, "U.S. Energy For the Rest of the Century,” EPRI Work-
shop Proceedings, Palo Alto, Calif., October 25-26, 1983.

Economic Report of the President, February 1982.

 

Nonresidential Buildings Energy Consumption Survey: Fuel Character-—
istics and Conservation Practices, DOE/EIA-0278, U.S. Department of
Energy, Washingtomn, DC, June 1981.

 

 

Nonresidential Buildings FEnergy Consumption Survey: Building Character-
istics, DOE/EIA-0246, U.S. Department of Energy, Washington, DC,
March 1981.

 

Nonresidential Buildings Energy Consumption Survey: 1979 Consumption
and Expenditures, Part 1l: Natural Gas and Electricity, DOE/EIA-0318/1,
U.S. Department of Energy, Washington, DC, March 1983,

 

 

1980 Annual Survey of Manufacturers: Fuels and Electric Energy Con-
sumed, Industry Groups and Industries, M80(AS)-4.1, Bureau of the
Census, U.S. Department of Commerce, Washington, DC, August 1982,

 

 

Monthly Energy Review, DOE/EIA-0035(93/08), U.S. Department of Energy,
Washington, DC, August 1983.

 

Survey of Current Business, 62(7), Bureau of Economic Analysis, U.S.
Department of Commerce, Washington, DC, July 1982.

Survey of Current Business, 63(7), Bureau of Economic Analyses, U.S.
Department of Commerce, Washington, DC, July 1983.

 

G. Samuels, D. P. Vogt, and D. M. Evans, Shifts in Product Mix Versus

 

Energy Intensity as Determinants of Energy Consumption in the Manufac-

 

turing Sector, presented at Electric Power Research Institute Workshop
on Forecasting Industrial Structural Change in the U.S.A., October 25-
26, 1983.

U.S. Industry Outlook 1977 with Projections to 1985, U.S. Department of
Commerce, Washington, DC, January 1977.

1979 U.S. Industrial Outlook with Projections to 1983 for 200 In-

 

dustries, Industry and Trade Administration, U.S. Department of Com—
merce, Washington, DC, January 1979.
B-6

1983 U.S. Industrial Outlook for 250 Industries with Projections for
1987, Bureau of Industrial Economies, U.S. Department of Commerce,
Washington, DC, January 1983.

Industrial Energy Use, OTA-E-198, U.S. Congress, Office of Technology
Assessment, June 1983.

 

C. C. Burwell, Glassmaking: A Case Study of the Form Value of Elec-
tricity Used in Manufacturing, ORAU/IEA-82-9(M), Institute for Energy
Analysis, Oak Ridge Associated Universities, Oak Ridge, TN, July 1982,

 

C. C. Burwell, Industrial Electrification: Current Trends, ORAU/IEA-83-
4(M), Institute for Energy Analysis, Oak Ridge Associated Universities,
Oak Ridge, TN, February 1983.

 

Electric Power Supply and Demand 1983-1992, North American Electric Re=-
liability Council, Princeton, NJ, July 1983.

 

Energy Projections to the Year 2010, DOE/PE-0029/2, U.S. Department of
Energy, Washington, DC, October 1983.

 

G. Samuels, Options for Electricity Use and Management during a
Petroleum Shortage, ORNL-5918, Oak Ridge National Laboratory, Oak Ridge,
TN, January 1983,

 

 

GADS—Generating Availability Data System, Equipment Availability Report
1972-1981, North American Electric Reliability Council, Princeton, NJ.

Monthly Energy Review, DOE/EIA-0035/80, U.S. Department of Energy,
Washington, DC, July 1980.

 

13th Annual Review of Overall Reliability and Adequacy of Bulk Power
Supply in the Electric Utility Systems of North America, North American
Electric Reliability Council, Princeton, NJ, August 1983.

 

 

S. Kichen and L. Pittel, "Utilities: Are the Good Times Over?," Forbes
(December 5, 1983).

“"Around the State Legislatures,” Modern Maturity (December 1983—January

1984).

 
in NPOVS.

APPENDIX C

DISCUSSION OF CONCEPTS NOT INCLUDED FOR ASSESSMENT

Many reactor concepts were proposed and considered for assessment

A list of those concepts that were not selected for detailed

assessment follows. The exclusion of concepts was based primarily on
the ground rules although other considerations contributed to the selec-
tion process.

LWR

APWR -

ABWR -

CNSS -

Explanations are included with each concept.

The Advanced PWR by Westinghouse is considered suffic-
iently developed to be available now; hence, there is no
merit in NPOVS assessment of the concept as a future
viable option. Furthermore, safety relies substantially
on conventional and engineered systems.

The Advanced BWR by General Electric is considered suffi-
ciently developed to be available now; hence, there is no
merit in NPOVS assessment of the concept as a future
viable option. Some of the Advanced BWR features are re-—
flected in the small BWR and, thus, are being considered
in NPOVS, Safety relies substantially on conventional
engineered systems.

The consolidated Nuclear Steam Supply System concept by
B&W is based on available technology and included 1little
emphasis on passive safety.

Steam~Cooled LWR - This "Schultz-Edlund” concept has no current ac-

W-NUPACK

tive vendor promoting it. As a result, it is judged that
the concept will not be available as a demonstrated option
by 2010.

600 - The small [600 MW(e)], barge-mounted plant offers
numerous cost advantages based on the maximum use of fac-—
tory quality fabrication, standardization, and modulariza-
tion. Westinghouse proposes marketing the plant with an
NRC final design approval so that utilities would face
primarily only the site suitability issues in licensing.
NUPACK will probably incorporate other design simplifica-
tion and advanced fuel cycle features of the APWR. Al-
though NUPACK relies significantly on passive safety it is
more traditional in its approach, primarily employing en-
gineered safety features.

CE-Realistic Alternative Reactor - This concept calls for a self-

pressurizing, single vessel, reactor-steam generator mod-
ule. It is similar in many ways to the CNSS, but uses
natural circulation for powered operation and does not
rely on the use of control rods or soluble poison for con-
trol during burnup. Pressure feedback is the control me-
chanism under powered operation. Design simplification
has been employed to limit the effects of many anticipated
transients and traditional design basis events for conven-
tional LWRs, but the ultimate safety response would still
rely on the intervention of engineered safety features.
LMR

CANDU - The Canadian heavy water reactors have served their domes-
tic needs well and have been deployed in several other
countries. Thus it is a wviable option, but there is no
U.S. sponsor aand the concept depends on engineered safety
features for decay heat removal, A principal rational
cost advantage derives from its use of natural uranium.
However, this advantage is lost when enrichment exists, as
in the United States. A smaller reactor, CANDU 300 has
been announced recently which is to have improved features
for safety and reliability, but it relies on engineered
safety systems and does not meet a passive safety cri-
teria.

Large Pool — This collective term applies to several concepts that
are being demonstrated in other countries and some con-
cepts studied in the United States. The concepts reviewed
have no active U.S. vendors promoting them and, hence, are
not considered available by 2010. However, the EPRI-COMO
program recently turned attention to a large pool design.

Large Loop — The large loop IMR concepts (other than the LSPB) have
no active proponent that would accomplish a demonstration
of the concept by 2010. These concepts are designed with
active, diverse, and redundant safety and do not emphasize
passive safety. The economic approach to the large 1loop
LMR is based on the need for the breeder and thus do not
meet the economic ground rule with present and near—-term
fuel prices.

W~Pool - The Westinghouse pool LMR concept was one of the con-
tenders for the DOE support of advanced concepts. Orig-
inally relying on an integrated fuel cycle with on-site
reprocessing and refabrication, it was later changed not
to require the integral fuel cycle. Not enough informa-
tion and detail have been available to NPOVS to include
this concept in the detailed assessment.

Hybrid - The Stone & Webster concept is based on two vessels, one
for the core and one for components, connected by pipes.
Not enough information is available to NPOVS to include
this concept in the detailed assessment.

Thermal IMR - The moderated core, cooled by liquid metal, has no
current sSponsor. It is judged as not available by the
year 2010, and not enough information is available for it
to be considered in an assessment. There is little infor-
mation about its present economic potential or its passive
safety features.

IFR ~ The Integral Fast Reactor, based on metallic fuel, inte-
grated pyrometallurgical reprocessing and on-site fabri-
cation, with the emphasis on metallic fuel, is promoted by
ANL. The reactor portion of the concept was not developed
in sufficient detail for assessment. The 1lack of an
active vendor contributed to the concept not being judged
GCR

MSR

HTR -

GCFR -

AGR -

MSR -

Other -

C-3

available for deployment by 2010. However, features of
this concept have been incorporated in the SAFR and PRISM
concepts that are included in this report. Also, an
analysis of the fuel cycle is presented in Appendix E.

This collective name applies to various versions of the
High-Temperature Gas-Cooled Reactors. 0f these, the
"Side-by~Side"™ prismatic fuel concept was chosen for
assessment. Other concepts were not examined in detail
because the side-by-side modular concept had been selected
for detailed study within the U.S. HTR Program. However,
experience from the pebble bed concept now operating in
two German demonstration wunits was wutilized in the
study. The 860-MWe large HTR has been included as an
appended reference since much of the HTR technology devel-
opment has been related to this concept and because it has
significant passive features, see Appendix F.

The Gas-Cooled Fast Reactor has no current active propo-
nent and hence is judged not to be available by the year
2010. Also, the available designs for a GCFR do not in-
corporate significant passive safety features.

This British designed and operated GCR has reached a point
of virtual standardization in the Heysham II and Torness
single—-cavity PCRV designs. These plants share the large
capital investment requirements of the large HTR but at a
lower power rating due to lower gas temperatures for the
carbon dioxide coolant. Therefore, competitive capital
costs in a U.S. market would be very doubtful. Recent
tests at Hinkley Point B have shown adequate passive cool-
ing of the pressurized core to the PCRV concrete without
damaging fuel or 1liner; however, the depressurized core
cooling does require forced convection. As at Fort St.
Vrain, liner cooling of the PCRV must be maintained to re-
tain any released fission products resulting from a de-
pressurized loss of normal heat sink.

All Molten Salt Reactor versions are excluded from de-
tailed assessment since having uno current active pro-
ponent, they cannot become available by 2010. Designs for
molten salt concepts date back many years. Passive safety
is not advertised, although many passive features are evi-
dent and some can be considered "inherent” to liquid fuel
systems. Economic estimates that were made are all obso-
lete and cannot be used for evaluating economic viability.

A few other concepts ("exotica") such as the fluidized bed
reactor were briefly considered and rejected for lack of
design information, lack of a sponsor, and insufficient
other information.
APPENDIX D

R&D GOALS AND SPECIFIC REQUIREMENTS FOR
LIQUID METAL REACTOR (LMR) CONCEPTS

Table D.l is a detailed presentation of R&D needs judged by the LMR
designers as essential or important to the success of their specific
power plant designs.l™3 Similar needs have been combined. The table
also indicates which R&D needs might apply to other reactor concepts and
provides justification for inclusion of each need. It should be
emphasized that Table D.l includes only those R&D tasks required to
complete a design to meet requirements and specifications.

Several challenges were identified for the IMR industry in the sec-
tion dealing with advantages and disadvantages of the concepts. Consid-
eration is given here to general R&D goals which could help meet these
challenges. However, to put this discussion in perspective, two
assertions are made and potential goals formulated. First, the IMR has
a long-term potential for breeding to extend fuel resources. Therefore,
one goal should be to maintain the capability to meet this challenge.
We also assert that the worldwide nuclear program will be sustained
through the NPOVS time frame, that a significant market for IMR
converters and/or breeders eventually will develop, and that U.S.
industry will seek a share of this market. Thus, a second goal should
be to sustain a competitive IMR industrial potential in the United
States for a significant range in growth rates of domestic and foreign
power needs. A competitive industry would have an adequate number of
properly trained technologists, up—-to-date and appropriate facilities,
and a competitive design to sell. These general R&D needs have been
organized in the form of a hierarchy in Table D.2 where these goals and
programs have been categorized within three major headings. This table,
though preliminary, provides a framework for evaluating the importance
of various R&D activities. For example, one could determine the rela-—
tive importance of the listed and augmented R&D tasks under the scenario
of low, modest, and high growth rates for the utility industry. Pre-
liminary assessments indicate that the list of R&D tasks is not sensi-—
tive to the schedule; only the relative importance of the R&D tasks was
scenario dependent,
Table D.l. Specific research and development needs identified for the LMR concept

 

LMR concepts

for which the

R&D needs was Other concepts

identified (*) for which the R&D

or applicable (X) may be applicable Justification for the R&D need

 

Research and Development (R&D)

needs identified by the designer Demonstrates Supports Increases Investor or
as essential or important LSPE SAFR PRISM LWR HTR low cost licensing public confidence

 

SAFETY-RELATED REQUIREMENTS
Advanced Core Design

Evaluate core features which can assure * * X X X
a benign core response to core disrup-
tive accident initiators,.

Demonstrate a low—cost and reliable * * X X X X
approach tor a self-actuated shut-

down system, Provide experimental

verifications of this concept needed

for licensing discussions.

Develop high temperature, wide range, * * X X X X X
fission channels and ion chambers for

power monitoring at in-vessel locations,

and high sensitivity source-range fission

channels for startup monitoring.

Perform experiments to demonstrate the * * * X X X X
effectiveness of B4C as an in-vessel

shield and verify shield design.

Perform detailed shielding and flux

calculations needed for the design.

Perform core critical experiments at * * * X X X
ZPPR to provide nuclear parameters and

detector requirements, and test loading

sequences for all cores considered.

Provide analytical verification ot X X * X X
benign response of the core to all
design basis accidents,
Table D.1l. Specific

research and development needs identified for the IMR concept (continued)

 

Research and Development (R&D)
needs identified by the designer
as essential or important

Perform seismic analysis and tests to
predict the response of the core and
reactor assembly to a seismic event.
Validate the code used through
experimental tests.

Collect and apply data on joint failure
probabilities of- key components for use
in reliability and risk assessment
calculations,

Develop methodologies for assigning
probabilities for accident sequences
associated with core disruptive
accidents. These probabilities are
to be used in Event Trees and PRA
studies for core responses and
structural responses.

Evaluate the reactor system and other
system responses to earthquakes to
determine seismic event categories,
evaluate safety systems reliabilities,
and provide inputs to PRA studies,

Develop necessary input and perform
analyses needed to quantify containment
response event trees for accident
sequences,

LMR concepts

for which the
R&D needs was
identified (%)

or applicable (X)

Other concepts
for which the R&D
may be applicable

Justification for the R&D need

 

LSPB SAFR PRISM

* X *

LWR HTR

Demonstrates
low cost

X

Supports
licensing

X

Increases Investor or

public confidence

X

¢ -
Table D.l.

Specific research and development needs identified for the IMR concept (continued)

 

LMR concepts

for which the
R&D needs was
identified (*)

or applicable (X)

Other concepts
for which the R&D
may be applicable

Justification for the R&D need

 

Research and Development (R&D)
needs identified by the designer
as essential or important LSPE SAFR PRISM

Develop and verify a 3-D coupled thermo- * X *
hydraulic, mechanical, neutronic, tran-

sient code used to support the design

of inherently safe reactor cores.

Develop a thermal-hydraulics core to

characterize temperature and flow fields

in large cores,

Provide friction and wear correlations *
to support innovative core holddown
designs.

Perform testing and analysis to quantify *
corrosion and tission product migration

and plateout in a sealed vessel without

cleanup systems.

Develop and validate an analysis code * X X
that can calculate deformations of

various core components due to creep and

swelling and calculate mechanical loads

thereby produced.

Shutdown Heat Removal

Perform experimental simulations of DRACS, * * *
RVACS, and RACS systems to evaluate their
passive design features, demonstrate their
operating principles, and optimize their
performance. Review 1984 tests of the
CRBRP NDHX system and understand uncertain-
ties in the performance of this system at
low air flow conditions. Perform 3-D
thermal-hydraulic analysis of the RVACs
performance. Perform tests of associated
flow control devices.

LWR HTR

Demonstrates
low cost

Supports
licensing

X

Increases Investor or

public confidence

X

b-Q
Table D.l. Specific research and development needs identified for the LfiR concept (continued)

 

LMR concepts

for which the
R&D needs was
identified (%)

or applicable (X)

Other concepts
for which the R&D
may be applicable

 

Research and Development (R&D)
needs identified by the designer
as essential or important LSPB SAFR PRISM

Determine and increase, if necessary, X * X
.the immunity of the decay heat removal
function to sodium fires.

Perform tests to verify design margins * X X
for creep of bellows in piping systems
at elevated temperature,

FUEL-RELATED REQUIREMENTS

Integral Fast Reactor
Design and evaluate the performance of X * *
a metal (U-Pu-Zr) core. Test fuel

assemblies in to demonstrate per-
formance for normal and cff-normal

conditions.

Evaluate and demonstrate the repro- X * %
cessing and refabrication of metal

fuel .

Validate safety claims associated X *

with metal fuel.
Long-life core

Perform extended burnup tests at * * *
FFTF of core materials, and blanket,

and control assemblies. Demonstrate

RBCB performance at EBR-II.

LWR HTR

Justification for the R&D need

Demonstrates Increases Investor or

public confidence

¢-ad
Table D,1.

Specific research and development needs identified

for the IMR concept (continued)

 

LMR concepts

for which the
R&D needs was
identified (*)

or applicable (X)

Other concepts
for which the R&D
may be applicable

Justification for the R&D need

 

Research and Development (R&D)
needs identified by the designer

as essential or important LSPE SAFR PRISM

LWR HTR

Demonstrates
low cost

Supports
licensing

Increases Investor or

public confidence

 

Perform characterization test of X * *
cladding and duct materials at

FFTF to determine irradiation effects

at prototypic temperatures.

Perform transient tests of high- *
burnup fuel pins to demonstrate

reliable performance under upset

condirions.

Automated Fuel Fabrication

Evaluate approaches and provide
conceptual design of fabrication
processes and equipment system
requirements.

SYSTEM—-AND COMPONENT-RELATED
REQUIREMENTS

Plant Experience

Utilize operating reactor experience * *
and data to evaluate shielding pre-

dictions, core performance predictions,

and flux monitor responses, and verify
under-sodium—-viewing device performance.

Establish and test methods to detect, * *
locate, and fix steam generator leaks.

Fabricate a prototypic detection and

location system.

X

X

9-a
Table D. lc

Specific research and development needs identified for the LMR concept (continued)

 

IMR concepts

for which the
R&D needs was
identified (*)

or applicable (X)

Other concepts
for which the R&D
may be applicable

Justification for the R&D need

 

Research and Development (R&D)
needs identified by the designer
as essential or important LSPB SAFR PRISM

Improve the design of conventional cold * X X
traps or develop new designs which are

more reliable, thereby improving plant

availability.

Perform analyses and testing associated *
with the PHTS siphon breaker to determine

its position, size, erosion/corrosion
resistance and reliability,

Advanced Plant Technology

Perform hydraulic tests, using a scale * * *
model, of the temperature and fluid flow

of the plenum, IHX, vessel wall, and

reactor vessel under normal power and

natural circulation conditions.

Evaluate various candidate materials as *
in-vessel insulation between the closure

head and sodium surface, with particular

attention given to French designs.

Study the effectiveness of redan *
as a thermal barrier and pressure seal.

Develop an approach for automating main- *
tenance functions for a multi-module
reactor site.

Investigate methods for validation and *
verification of software used in reactor
control and protection systems.

LWR HTR

Demonstrates
low cost

X

Supports
licensing

Increases Investor or

public confidence

[~a
Table D.1.

Specific research and development needs identified

for the ILMR concept {(continued)

 

LMR concepts

for which the
R&D needs was
identified (*)

or applicable (X)

Other concepts
for which the R&D
may be applicable

Justification for the R&D need

 

Research and Development (R&D)
needs identified by the designer

as essential or important LSPB  SAFR PRISM

LWR HTR

 

Demonstrates
low cost

Supports
licensing

Increases Investor or

public confidence

 

Investigate advanced instrumentation and * * *
control stress autcmation, distributed

control multiplexing, improve measurement

sensors and systems, simplified maintenance,

use of artificial intelligence, operator

aids, and human engineering.

Steam Generator Performance

Conduct steam generator endurance tests X * *
to demonstrate long-term integrity.

Investigate other options to simplify

the overail system, Investigate the

performance of booster tubes in steam

generators. Recommend or reference

system and ideatify necessary key

features tests.

Verify that existing inspection tech- * X X
niques for S$SGs meet code requirements

and develop new techniques which might

be used at elevated temperatures and

in the presence of sodium.

Improve computer code predictions of * X X
5G performance under low sodium flow
conditions.

Improved Materials
Obtain code approval for advanced * * *

materials and simplify or improve
code rules for conventionial materials.

X

X

X

8-
Table D.l.

Specific research and development needs identified

for the IMR concept (continued)

 

ILMR concepts

for which the

R&D needs was Other concepts
identified (*) for which the R&D
or applicable (X) may be applicable

Research and Development (R&D)
needs identified by the designer
as essential or important

Perform thermal striping tests for
various candidate materials for
upper internal designs.

Evaluate materials which could
enhance radiative heat transfer
associated with decay heat
removal systems.

Evaluate purification methods for
primary sodium and cover gas in a
sealed, vessel during normal operation,
and during refueling.

Verify the capability of an under
sodium viewing system to satisfy
in-service inspection requirements
and refueling inspection require-
ments.

Perform component testing and obtain
information from the British and
French concerning location of

failed fuel using a sodium sipper.

Develop a high sensitivity, fission
channel with remote signal trans-
mission capabilities for use as a
monitor of initial core loadings.

Justification for the R&D need

 

LSPB SAFR PRISM

* *

LWR

HTR

Demonstrates
low cost

X

Supports
licensing

X

Increases Investor or
public confidence

6—d
Table D.1l.

Specific research and development needs identified for the LMR concept (continued)

 

Research and Development (R&D)
needs identified by the designer
as essential or important

LMR concepts

for which the
R&D needs was
identified (*)

or applicable (X)

Other concepts
for which the R&D

may be applicable Justification for the R&D need

 

LSPB SAFR PRISM

 

Advanced Sodium Component Feature Tests

Develop advanced pool-pumps such as
a compact, self-cooled electro-—
magnetic pump and a shrouded inducer
pump.

Test contreol rod drive-line designs
which provide inherent negative
reactivity in response to core
excursiaons.

Increase confidence in the use of
flexible joints in piping system
through their testing at EBR-II.

Determine flow distributions and
investigate vibrations for IHXs
through tests of physical models.

Develop a conceptual design of an
innovative refueling and main-
tenance system, and demonstrate
key features by testing.

Demonstrate the functioning of
the core support systems through
tests using an engineering scale
model.,

Develop methods for under-sodium,
in-service inspection of heat
exchanges.

X * %
X X *
X * X
X * X
*
*
*

Demonstrates Supports Increases Investor or
LWR HTR low cost licensing public confidence
X
X X
X X X
X

01
Table D.l.

Specific research and development needs identified for the IMR concept (continued)

 

Research and Development (R&D)
needs identified by the designer

as esseuntial or important

LMR concepts

for which the

R&D needs was Other concepts
identified (*) for which the R&D
or applicable (X) may be applicable

Justification for the R&D need

 

 

Demonstrates Supports

LSPB SAFR PRISM LWR HTR low cost licensing

Increases Investor or
public confidence

 

Perform tests of the reliability * X X
of the bearings and seals for the
rotating plug of the closure head.

Perform test to demonstrate the * X *

performance under design and
abnermal conditions of centri-

fugal pumps, inducer pumps, and

electromagnetic pumps.

Perform tests to verify that
primary and secondary control
rod systems satisfy design
requirements.

X

 

*One or more of the needs of the associated list was specifically identified for this design.

XOne or more of the needs

of the associated list would be applicable for this design.

11-C
Table D.2. A hierarchy of R&D tasks to keep the IMR/Breeder option as healthy and competitive
as possible considering a realistic range in future nuclear energy usage

 

Maintain an adequate work force of
technologists and appropriate
up-to-date facilities

Continue to improve IMR designs
so that the concepts available
will be competitive

Continue to solve institutional problems
and improve the marketability of IMR
concepts

 

Support R&D that increases the design
options available for new LMR concepts
and significantly improves the
technology

Materials research for higher
operating temperatures and
improved efficiencies

Steam generator designs to
eliminate sodium—water reactions
(double-wall concepts) or provide
instrumentation for more accurate
and reliable detection

Improved instrumentation and
control to incorporate advances
in automation, artificial
intelligence, digital control
etc.

Advanced oxide fuel designs for
higher burnup, and metal fuels
for safety and reprocessing
advantages

Improved core designs for a
once-through cycle so that
reprocessing is not necessary
for cost competitiveness

Support university research to maintain
a continuous supply of technologists

Complete and demonstrate technical
solutions to long-established
design challenges

Demonstrate passive safety against
core~disruptive accidents

Establish the plant size and con-
figurations which have the potential
for lowest power costs

Demonstrate cost competitive off-
site and/or on-site reprocessing
and refabrication

Support standardization of design
for improved licensability

Demonstrate simpler, passive, decay
heat removal concepts

Strive for a significantly better
reactor design with convincing
advantages in cost, public acceptance,
licensability, etc.

Produce, test, and qualify whole plant
designs and/or components such as
steam generators, pumps, etc., which
can be sold to non-U.S. markets

Provide monetary incentives for
utilities and industry to build LMR
demonstration plants and facilities
(for example, license them as R&D
facilities

Encourage and support R&D that increases
consumption of electrical energy within
the guidelines of national policy (for
example, support storage battery research
to make electric cars attractive)

Decrease the complexity and shorten the
time required for licensing (for example,
one step licensing process)

Increase utility involvement in IMR
technology through personnel exchanges,
joint research, etc.

Complement non-U.S. R&D and commerciali-
zation activities so information exchange
with other countires will be mutually
beneficial

If required, obtain support for LMR tech-
nologists and R&D facilities from closely
related areas such as defense, space
research, etc., so that their skills will
be maintained

 

c1-a
REFERENCES FOR APPENDIX D

1.

3.

LSPB Research and Development Requirements, CDS 500-6, U.S. Depart-—
ment of Energy, Washington, D.C., September 1984. APPLIED

TECHNOLOGY.

SAFR Requirements for Base Technology Program, 149T1000002, Rockwell
International Rocketdyne Division, Canoga Park, California, January
1985. APPLIED TECHNOLOGY.

Letter number XIL-897-850016 from L. N. Salerno to Francis X
Gavigan, dated January 11, 1985, "WBS2BO.5 — Initial PRISM R&D
Requirements Statements.,”
APPENDIX E

LIQUID METAL REACTOR (LMR) FUEL REPROCESSING-
REFABRICATION EVALUATION

Je Te Bell and D. C. Hampson

This evaluation was developed to compare IMR fuel recycle systems
for oxide and metal fuels. An adequate data base was not then avail-
able, partiecularly for the metal fuel of the Integral Fast Reactor
(IFR). An extensive program of study is now in progress at Argonne
National Laboratory to develop a metal fuel system for the IMR and thus
Fill this wvoid. Although the following evaluation is preliminary, it
illustrates the questions that must be resolved to arrive at a final
comparigon of the fuel systems. The conclusions should be viewed quali-
tatively since the quantitative results are subject to revision as new
data are developed by the Argomne study. However, our principal con-
cerns for Argonne's estimates of the amount of research and development
required and for the project costs, both of which are lower than our
analysis indicates, have not been alleviated by work published to
date. On the other hand, the scientific quality of the process research
reported appears to be excellent.

This evaluation classifies fast reactor fuels as either oxide or
metal. Reprocessing of oxide fuels is considered only with the Purex
process, and the results are based primarily on ORNL experience over
three decades. Metal fuels reprocessing is considered for an Argonne
National Laboratory (ANL)-developed pyroprocess that includes molten
salt and electrochemical techniques. The discussion of Purex processing
will relate directly to any fast reactor concept with mixed-oxide (MOX)
fuel, while the discussion of metal fuel reprocessing relates directly
to the Argonne Integral Fast Reactor (IFR). The total reactor output
for each concept is assumed to be 1300 MW(e). Most studies and programs
for oxide fuel reprocessing have been for substantially larger plants.
The small size here is chosen to match the IFR concept. These two fuel
reprocessing schemes will be compared, and the resulting analysis should
be generically applicable to other reactor concepts when metal and oxide
fuels are considered.

The metal fuel could be processed by the Purex route with minor mo-
difications. However, this would discard one of the prime benefits of
the pyroprocess, which is that the fuel remains essentially in a metal
state which is amenable to refabrication steps developed for the metal
fuels. The metal fuel refabrication process is somewhat less compli-
cated than the pellet pressing process envisioned for the MOX fuels.

A cost estimate for the Purex reprocessing of oxide fuels will be
more accurate than that for metal fuels because the Purex process, in-
cluding management of its wastes, has already been developed to the con-
ceptual design stages for the Hot Engineering Facility (HEF) in 1978 and
the Breeder Reprocessing Engineering Test (BRET) in 1984. The fused-
salt electro-refining process (FSER) proposed for reprocessing metal
E-2

fuels is in the development and proof-of-principle stages. Although the
FSER process is less developed, our analysis will assume that this pro-
cess for metal fuels is valid, in principle, and that design of equip-
ment for a Hot Experiment Plant (HEP) could begin in 1986 for the oxide
fuel and in 1989 for the metal fuel. It is assumed that an HEP for
either process 1is required to provide design data for a commercial
demonstration reprocessing plant.

This evaluation applies to the 1985-2005 time frame and is based on
a commercial demonstration reactor in 2005. The schedule would require
a fuel reprocessing demonstration plant about 3 years later. To meet
the 2005 goal, we have assumed that certain major facilities are avail-
able now. The Fast Flux Test Facility (FFTF) at Hanford would be used
for irradiating oxide fuels, and the Fuels and Materials Examination
Facility (FMEF), also at Hanford, would be used for an oxide fuel HEP.
The Experimental Breeder Reactor No. 2 (EBR II) and associated Fuel
Cycle Facility (FCF) at Idaho Falls would be used for irradiating metal
fuels and for a molten salt-electrochemical HEP, respectively.

In both cases, it is assumed that the existing facilities (FMEF and
FCF) can be modified and equipped to provide the functions of the HEP.
Each HEP would be used to develop and demonstrate the proof-of-principle
of the respective process and would receive irradiated fuels supplied by
the associated reactor (FFTF for MOX, and EBR 11 for metal fuels). It
should be noted that the FFTF does not have blanket elements, which
would be present in a demonstration fast breeder reactor (FBR). The
proof—-of-principle demonstrations would provide the technical informa-
tion necessary for the design of a demonstration facility. It 1is
assumed that these HEPs would contribute sufficiently to design infor-
mation to justify a second demonstration plant. However, the latter may
be a first-of-a-kind commercial plant. Without these facilities, it is
unlikely that either process could be commercially demonstrated in the
NPOVS time frame,

E.l1 Schedules for Development of Commercial Demonstration

The 2000—2010 period has been selected as a feasible objective for
demonstration of a selected new power reactor and the associated fuel
cycle. Although the necessary time for reprocessing would be 3 to
5 years after the demonstration reactor goes on line, a fully developed
fuel cycle would be essential to adoption of the IFR concept.

To establish schedules for developing the reprocessing/refabrica-
tion systems for oxide and metal fuels, it was presumed that an existing
facility could be modified for specific needs of each process. This
further implies that the use of the FCF for metal fuels or the FMEF for
oxide fuels would be adequate to provide proof-of-principle information
for either fuel cycle. However, use of these existing facilities would
not provide the hard-schedule financial data for construction that are
required for the commercial phase; such data would be a product of the
demonstration phase.
E-3

The major difference between the two schedules is that the metal
fuel reprocessing must be initiated with a process/waste development
phase, while this work is not needed for the oxide fuel reprocessing
program. Adequate development work has been done on the oxide program
to permit immediate initiation of the design activities. The metal fuel
recycle program would be divided into three components:

® process/waste development
® proof-of-principle runs in the FCF, and

® design and construction of a demonstration fuel cycle plant at a
location to be determined.

The oxide fuel recycle program would be divided into two major compo-
nents:

® proof-of-principle rumns in the FMEF, and

® design and construction of a demonstration fuel cycle plant at a
location to be determined.

Three constraints have been incorporated into the schedule for
metal fuel processing:

1. The process must be proven on a laboratory scale prior to design of
the modifications for the FCF.

2. Cold-testing of the process must be successfully completed in the
FCF prior to start of design for the demonstration plant.

3. Hot-testing of both the separation process and the waste process
should be completed prior to start of construction of the demon-
stration plant.

The one constraint that was deemed necessary for oxide fuel processing
was that the integrated hot-testing should be completed prior to start
of construction of the demonstration plant. Refabrication of both fuels
must be included in the respective HEP, and proof-testing should be com-
pleted before initial construction of the demonstration facility.

The two schedules are based on a commitment to conduct the various
activities within the time frames shown. This is critical, and delays
in any phases of the program would be reflected in corresponding delays
in subsequent phases. As can be seen in the table at the end of this
section, there 1is little room for slippage 1in either fuel cycle
schedule. These are the most optimistic feasible schedules and are
given only to show that a 2005 objective could be achieved. These
schedules can be accomplished only if the following are done:
E-4

l. The metal-fuel schedule is started immediately.

2. Experimental reactor space for fuel testing is dedicated to this
effort.

3. Existing facilities are available and dedicated to the HEPs.
(There is no time for construction of new HEPs.)

4. Waste management 1is developed parallel to the chemical pro-
cessing. (This will require an additional effort in the metal
fuels program since little work has been done for this waste to
date. The necessary development may extend past the 1997 date.)

5. Fuel fabrication must be developed simultaneously with the chemical
reprocessing.

6. Plutonium must be available for fuel testing and for irradiation in
order to have adequate spent fuel available for either HEP in the

1991 time frame.

Summary Schedule for Development

 

Metal Oxide

Laboratory experimentation started 1985 NR*
Complete laboratory experimentation 1987 NR*
Start equipment design for FCF or FMEF 1988 1988
Start modifications to FCF or FMEF 1989 1989
Install equipment in FCF or FMEF 1990 1991
Complete base experimental program 1997 1993
Start design demonstration plant 1994 1989
Start construction of demonstration plant 1998 1993
Start design of demonstration equipment 1996 1991
Install demonstration equipment 2000 1996
Complete construction 2002 1997
Start operating the demonstration plant 2004 1999

 

*NR indicates steps not required. Thus, the oxide schedule could
be advanced one year.

E.2 The Two Processes

The flowsheet for pyrochemical processing of metal fuell and an
equivalent Purex flowsheet for oxide fuels? were evaluated. The block-
diagram flowsheets described below are based on the conceptual design
efforts for commercial~scale fuel cycle facilities to serve 1200 to
1400 MW(e) electricity generating capacity (nine PRISM modules or four
SAFR power paks). Refabrication flowsheets follow Argonne information
for metal fuels and the mechanical-blending pellet forming process for
oxide fuels.
E-5

E.2.1 The Pyrochemical Process for Metal Fuels

The present IFR concept (1985) includes a reactor-generator system
operating in a fuel break-even converter mode but constructed with the
potential of operating in a fuel breeder mode. For this analysis, mul-
tiple reactors are considered in a 1300-MW(e) unit with associated fuel
cycle facilities. A fuel burnup of 117 is proposed without an axial
blanket. The break-even operation requires a reprocessing system with a
throughput of 7.5 tonnes/year (t/a) (30 kg/d) of core fuel and 8 t/a
(33 kg/d) of combined internal and radial blanket fuel.3 However, if
the reactor has potential for future fuel breeding, the reprocessing
facility must be initially constructed to accommodate the breeder sys-

tem. The previously proposed IFR breeder required reprocessing
capacities of 7.5 and 18 t/a (30 and 72 kg/d) of core and blanket fuel,
respectively. This analysis will consider only the higher—-capacity

requirements for the breeder mode. Again, no axial blanket 1is assumed.

A block-diagram flowsheet for reprocessing metal fuel is shown in
Fig. E.l. A special committee appointed by the University of Chicago
has reviewed the chemistry proposed for the reprocessing of metal fuel;
further discussion is not presented here. The core and the blanket fuel
initially must be processed separately, and the plutonium from the
blanket fuel is added to the core fuel. The blanket fuel is dis-
assembled, chopped, and dissolved from the cladding by an electrodis-
solution process. The fuel dissolved in cadmium is contacted with a
salt mixture to oxidize the plutonium to PuCl3, which transfers into the
salt solution. Some of the uranium and most of the fission products
will also be oxidized and dissolved into the salt. Noble metal fission
products will remain predominantly with the unoxidized uranium. After
this halide slagging, the salt solution is transferred to the process
line for the core fuel and 1is coprocessed through the electrolysis
scheme. Some problems with the halide slagging process include waste
handling and control of the metal oxidation. The plutonium is oxidized
by adding UCl3 to the salt solution. The amount required will depend on
the amount of fission products that oxidize before the plutonium
oxidizes. The added uranium also replaces that consumed in the blanket.

As shown in block form in Fig. E.l1, the fuel bundles are dis-
assembled and the individual core fuel pins are punctured to release
gaseous fission products and then chopped into short segments. If an
axial blanket had been included, that material would be mechanically
separated and transferred to the decladding and the halide slagging pro-
cess for blanket fuel. The chopped core fuel is dissolved into cadmium;
the cladding does not dissolve. The salt solution from the halide
slagging of the blanket fuel is added to the cadmium. This results in a
two-phase liquid system — a molten salt phase in contact with a molten
metal phase. Also, the cladding and the undissolved fuel make up a
solid phase in the cadmium liquid metal. 1In electrorefining, the metal
phase is used as the anode, and a cathode is placed in the salt phase.
The PuCl3 is electro-reduced to plutonium from the salt phase onto the
cathode. Plutonium is transferred from the metal to the salt phase by
oxidation at the anode to PuCls. The same electrolysis mechanism
E-6

ORNL DWG 85-492

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

CaCl,

MgCl,

BaCl, SOLIDS *BeD CRUCIBLE

3 WASTE" SKULLS
CLAD DECLADDING 4 5.6 9
ME
waste [+ stee . stAGemg [ e accounTaeruT |
SALT

 

 

 

 

 

 

 

 

 

 

 

2 10
1
WASTE @EMBLE FUME | o ACCOUNTABILITY i

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

  

 

 

   
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

METAL(s) TRAP
BLANKET
STORAGE FUME FUEL
TRAP Al FABRICATION
BLANKET | | SODIUM | WATER
FUEL REMOVAL
CAUSTIC a
EVAPORATOR
COMPRESSED
GAS, STORAGE
CORE SODIUM LOW- o
— LEVEL FUME
FUEL REMOVAL WASTE 7 TRAP
N
soLios 1] CORE FUEL
WASTE FABRICATION
FUME
| TRAP A} —w
STORAGE SAMPLE
ANALYSIS
2
WASTE : 22
METAL () DISASSEMBLE FEED .._-g
PREPARATION r
FOR
20
— 1 \2 FUME FABRICATION .
AL REMOVE TRAP
WASTE "] pLENUM - cd STORAGE
Ciz 24
15,16 ¥ . PRODUCT 21
13 DISSOLUTION ELECTRO- CONSOLIDATION ENCAPSU-
SHEAR CADMIUM REFINING b 23 N
REMOVE T 22 19
c AND
o
14 ° 7 EL%E'ST?:(-;I-)‘E wetaL | | VITRIFICATION
WASTE
g L e | | REmOVE PRODUCT CONSOLI~
BASKET DATION
SALT
l ;' CONVERSION
TO OXIDE

 

 

 

 

Figure E.l. The block diagram flowsheet for pyro—electro—-chemical
reprocessing fast reactor metal fuel. Numbered blocks indicate space
requirements in the containment, and multiple numbers per block indicate
duplicate equipment. The decladding step, block 3, has not been defined
but is thought to be feasible by an electrochemical process. Any such
process will likely require at least three additional space require-
ments. The letters a, b, ¢, d indicate a sequence of steps following
the electro-refining. During startup of the reactor with enriched
uranium, the core uranium cannot be isotopically diluted with blanket
uranium, and the blanket fuel may need to pass two times (or more)
through the halide slagging step to reduce to uranium content.
E~7

applies to the reduction of UCl3 to uranium and of ZrCly to zirconium,
and the product is a mixture or an alloy of the three metals. Control
of the reducing potential can partially segregate the metals on the
cathode.

E.2.2 The Purex Process for Oxide Fuels

The assumed LMR design, as for the metal fueled system, is a
1300-MW(e) facility. The core and blanket fuels can be processed in the
same equipment. A reprocessing facility on the site for this 1300-MW(e)
complex would require a capacity of 35 t/a (8 t/a of core and 27 t/a of
blanket fuel) or about 140 kg/d at 250 d/a.

The Purex process for reprocessing oxide fuels 1is diagrammed in
Fig. E.2 and has been previously described.? Both the core and the
blanket fuels are processed through the same equipment. The head-end
operation involves disassembly of the fuel bundles and chopping of fuel
into 1-in. segments. Dissolution includes transfer of fuel from the
cladding into nitric acid and removal of cladding hulls from the pro-
cess. In the feed preparation step, the chemistry of the dissolved fuel
is adjusted for solvent extraction.

The details of solvent extraction strongly depend on the character-
istics required for the plutonium and uranium products. OQur evaluation
will be based on a conventional Purex solvent extraction, in which the
solvent extraction step yields pure uranium and plutonium products. The
product-conversion step converts aqueous nitrate solutions of plutonium
and uranium into the respective oxides. The reference design includes
oxalate precipitation followed by thermal decomposition.

E.2.3 Fabrication of Metal Fuels for IFR

Fabrication of the metal core and blanket fuels will involve
separate operations in the initial steps. However, the same equipment
can be used during and after the fuel-rod assembly stage. All fabrica-
tion steps must be conducted remotely within thick containment shielding
because both the recycle uranium from the halide slagging and the repro-
cessed plutonium/uranium/zirconium will retain appreciable quantities of
fission products. Figure E.3 shows a block diagram for metal fuel
fabrication, which is in accord with the fabrication of EBR II fuel. As
in the reprocessing, the capacity for fuel fabrication must be capable
of quantities for the breeding design basis, so that the IFR will have
breeding as a future option.

Batch sizes may be 1limited by criticality considerations. The
criticality safety analysis for these process steps has not been com-
pleted, and therefore the effect of criticality on batch sizes cannot be
firmly evaluated. IFR breeder systems with a total output of 1300 MW(e)
will require a total processing-refabrication capacity of 25.5 t/a
(30 kg/d and 72 kg/d for the core and blanket fuels, respectively, for
ORNL DWG B85-493

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

   
   

 

 

 

 

 

   

 

 

   

 

 

 

 

  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

LOW--LEVEL
LIQUID WASTE
IRRADIATED | Na
FUEL REMOVAL TO HEPA
FILTERS
AND STACK
| FUEL STORAGE |
1" 1
NO. 1 SOLVENT NO. 2 SOLVENT
1 I0DINE CLEANUP TO Law CLEANUP [~ TO LAW
CROPPING ' : Ty P
1
» CONDENSERS Ru WASTE _ 9.10 L g2 12 N : I ! 12 3
Ru TRAP :
ACCOUNTABILITY st CYCLE ||
DIGESTION 2nd U CYCLE
SHEARING —  AND FEED CODECON— 3rd U CYCLE H—={U0,(NO3)
DISSOLUTION , TANKS CLARIFICATION TAMINATION U-Pu PARTITION : ; e
DISSOLVER | 14
5 RESIDUES ,
HULLS MONITOR 25 ! | P7] 2nd Pu CYCLE
EXCESS WATER TO —TO VOG ‘
3 LOW-LEVEL WASTE T0 [ 14
METAL r“‘“' 4" 3rd pu cveLE
WASTE | RECYCLE WATER ‘24
I 22 TO PROCESS CONDENSER g 15,16
ACID RECYCLE ACID |23 % PLUTONIUM
RECOVERY l '| TO PROCESS ] £ K Y LE NITRATE
STORAGE
21 ——=T0 VOG J -
LAW —— LAW VITRIFICATION -
(WASTE} CONCENTRATION OFF—GAS , [~ Rv WASTE Pu FILTRATE PLUTONIUM
PRECIPITATION
CONDENSER 1 18
26 Pu DISTILLATE | pLuTONIUM - TO FUEL
HAW CALCINATION FABRICATION
CONCENTRATION
—— SOLIDS, AQUEOUS LIQUIDS, 27 l 28
AND GASES
HLW VITRIFIED
——— ORGANIC LIQUIDS VITRIFICATION HLW

 

 

Figure E.2. The block diagram flowsheet for Purex reprocessing of fast reactor oxide fuel.
Numbered blocks indicate process steps that require space in the containment facility, and a number
common to more than one block indicates that these steps share a space. For example, the solvent
extraction steps, #9, are accomplished in an eight-pack of centrifugal contactors.

8-
 

E-9

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

ORNL DWG 83-523

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

6 7 20
CAN WELD AND STORE
LOADING LEAK TEST
19
S 8 STRAIGHTNESS
INSPECT AND BOW TEST
4 9 19 1
DEMOLD BOND TENSILE
AND SHEAR TEST TEST
h.2.3 I 9
BLANKET INJECTION STRAIGHTEN
FUEL CASTING SPADE
21 L‘19 19
WASTE HEXAGONAL
[—- ASSEMBLY CAN INSERT
CORE INJECTION STRAIGHTEN
FUEL CASTING SPADE
13 110, 11,12 18 1
DEMOLD AND BOND
SHEAR TEST
14 17
INSPECT BOND
15 16
CAN WELD AND
LOADING LEAK TEST

 

 

 

 

 

 

 

Figure E.3. The block diagram flowsheet for fabrication of fast
reactor metal fuel. Numbered blocks indicate processes or steps that
require space in the containment facility, and multiple numbers per
block indicate duplicate equipment. A single number on several blocks
indicates that these steps are accomplished in a common space.
E-10

250 d/a). Thus, as a result of criticality considerations, the core and
the blanket process may each require more than one injection casting
unit. The cast fuel will be cooled and the mold removed. The spent
mold becomes TRU waste but could, perhaps, be cleaned to low-level
status. The fuel will then be inspected and a chemical analysis com-
pleted. The fuel that passes inspection will be loaded, along with some
sodium, into new cladding. The cladding will be subsequently welded and
leak-tested. Approved fuel rods will be bonded into bundles, and this
bonding will be inspected. The positioning spade on the bottom of the
fuel rods will be straightened, and the rods will be loaded into an
assembly. Finally, each assembly will be equipped with the hexagonal
can 1insert and welded. Tensile testing will be performed, and the
approved assemblies will be examined to assure meeting the requirements
for insertion into the reactor.

E.2.4 Fabrication of Oxide Fuels

Feed material for fabrication of oxide fuels will have relatively
low gamma activity because the Purex process removes essentially all of
the fission-product elements from the plutonium/uranium products.
Therefore, this fabrication can be housed in a low-level containment
facility. However, automated remote operations and maintenance will be
required because of the long-term increase in the even—-numbered pluton-
ium isotopes. The 1300-MW(e) reactor system will require 8, 9, and
18 t/a of core, axial, aund radial blanket fuels, respectively. While
the Purex reprocessing must handle all 35 t/a of spent fuel, the fabri-
cation facility will prepare only the core fuel (8 t/a); the blanket
fuel (27 t/a) will be purchased from an independent vendor. The vendor
could be any of those that currently supply LWR fuel.

A block-type schematic for fabrication of oxide breeder fuel is
shown in Fig. E.4. This flowsheet is in accord with the fabrication
flow-sheet for the Secure Automated Fabrication line." The recycled
Pu02 and the U02 are first blended in the proper batch quantities. The
mixed oxide is milled to achieve uniformity, and a binder material is
added. The fine material is then compacted into granules, and a lubri-~-
cant 1is added. This mixture is pressed into pellets and loaded into
boats for high-temperature treatment, which serves to remove the binding
material and sinter the pellets. The sintered pellets undergo several
grinding, gauging, cleaning, and inspection steps before they are loaded
into the cladding jackets. The loaded jackets are filled with helium
and welded closed. Helium leak-testing and ultrasonic testing are car-
ried out successively. The fuel rods proceed through fissile assay and
physical inspection, and rods that pass all inspections are wrapped with
spacer wire and assembled into bundles. A final imnspection of the
assembled bundles completes the fabrication process.
E-11

ORNL DWG 85-524

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1 2 3 4q
uo, - MILLING COMPACTION LUBRICANT
Pu0, ———={ BATCHING AND AND ADDITION
RECYCLE —e BLENDING BINDER ADDITION GRANULATION
11 10 9 7,8 5,6
PELLET PRESSING
INSPECTION STORAGE AND SINTERING LOADING
2 | i i
CLi‘m')'”G BOAT CLEANUP CHEMICAL
GAGING AND INSPECTION ANALYSIS
13 14 15 16 17
COLUMN MAKEUP DECON- HELIUM FILL HELIUM LT i
PIN LOAD TAMINATION AND WELD LEAK TEST TESTING
26 25 22,23,24 20, 21 18, 19
ASSEMBLE VISUAL FISSILE ASSAY
STORE AND WIRE DIMENSION COMPONENT
INSPECT WRAP CHECK PLACEMENT

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Figure E.4. The block diagram flowsheet for fabrication of fast
reactor oxide fuel. Numbered blocks indicate processes or steps that
require space in the containment facility, and multiple numbers per
block indicate duplicate equipment.
E-12

E.3 COSTS OF DEMONSTRATION PLANTS FOR REPROCESSING
AND REFABRICATING METAL AND OXIDE FUELS

Costs have been estimated for the fuel cycle plants for fast reac-
tor facilities with 1300-MW(e) outputs. In Table E.l1, through-puts for
such reprocessing facilities are given as tonnes per year (t/a) for
metal fuel from an IFR conceptual reactor operating as a breeder and for
oxide fuel from a PRISM conceptual breeder reactor. Reprocessing of
either type of fuel is conducted in remotely operated, shielded facil-
ities. Refabrication of both the core and blanket of the metal fuel is
conducted in remotely operated, shielded facilities. However for oxide
fuels, only the core is refabricated remotely, the blanket fuel is pur-
chased from a commercial vendor.

Table E.l. Fuel throughputs (t/a) for reprocessing and
refabricating metal and oxide fuels from
1300-MW(e) fast reactors

 

Total for
Refabri- remote
Core Axial Radial Reprocess cation Purchase operations

 

Metal fuel 7.5 0 18 25.5 25.5 0 51
Oxide fuel 8 9 18 35 8 27 43

 

The costs of reprocessing and refabrication facilities are highly
dependent on the sizes of the containment buildings, because of the
shielding and ventilation requirements. Therefore, the relative costs
of fuel cycles for oxide or metal FBR fuels can be based on the relative
sizes of the required facilities. As a first approximation for compar-
ing costs, we have assumed that capital costs are proportional to the
number of process steps that require space in the shielded, remotely
operated facility. We have numbered such steps in Figs. E.l—E.4.

Several studies have calculated and examined the effects of various
parameters on reprocessing costs for oxide fuels. Those are summarized
in Ref. 5. Those costs and the costs herein are expressed as current-
dollar levelized costs (1984 dollars). If the annual levelized fixed
charge rate had been based upon constant dollars, the apparent cost
would have been lower. Using the same scaling factors (0.5) that are
used in those studies and assuming, as did Delene et al.,5 that a facil-
ity for reprocessing 150 t/a would cost $1.02 billion, we estimate that
the capital cost of an oxide fuel reprocessing facility for a 1300-MW(e)
LMR would be $492 million. The validity of this cost figure is sup-
ported by the recently published costs for the newest LWR reprocessing
plant in the Federal Republic of Germany.® 1In that document, $1.6 bil-
lion is the cost reported for a 500-t/a LWR reprocessing plant (exclud-
ing costs of the refabrication facility). When scaled downward to the
35-t/a plant, this yields $423 million. Since the costs of reprocessing
LMR fuels are generally higher (up to 50%) than those for IWR fuels, the
E~-13

two derived costs are comparable, and confidence in their wvalidity is
increased. However, preliminary results of an ORNL study currently in
progress for a small onsite facility suggest a cost of $270 million com-
pared to the above derived $492 million for reprocessing of oxide fuels
in a 35 t/a facility.7 We have used these two cost figures to calculate
lower and upper boundary costs for reprocessing and refabrication. The
oxide fuel reprocessing costs were derived from these values, while the
costs for refabricating the oxide fuel and blanket were based on the
high end of the values given in Table 2.12 of Ref. 5. The high-end
values were used to reflect the cost disadvantage for a small-throughput
plant.

The corresponding cost estimates for a metal fuel reprocessing and
refabricating plant have not been prepared in the same detail as those
for the oxide fuel plants. Because of this, the values derived for the
metal fuel cycle do not have the same validity as those for the oxide
fuel cycle. However, we think the results are generally correct and
acceptable as a basis for comparison. A more detailed design study
would be required to provide cost estimates of equal validity to those
for the oxide fuel plants.*

Since the number of major steps required for processing of metal
fuels is comparable to that required for oxide fuels (see Figs. E.l1 and
E.2), the cost of the oxide reprocessing plant was used as a cost basis
for the metal reprocessing plant. However, metal fuel reprocessing
equipment may not require the head height that is needed for oxide
reprocessing equipment. Therefore, the initial base cost for the metal
fuel was reduced by 20% before scaling for capacities was done. The 207%
reduction is based on the resulting reduced need for concrete and rein-
forcement, relative to the total reinforced concrete requirements for
the total cell. This reduced base cost was then adjusted for size based
on capacity, using a 0.5 scaling factor.

The facility requirements for reprocessing or refabricating metal
fuels are similar. Therefore, the costs for refabrication of the metal
fuels were based on the relative number of steps required for refabrica-
tion (i.e., 20) as compared with the number of steps required for
reprocessing of the metal fuel (i.e., 25) to the 0.5 power. Thus, the
refabrication capital costs are: (20/25)%9+5 = 0.894 times the metal
reprocessing capital costs.

The total waste costs were based on the rate of 1 mill/kWh and
hence were $9.1 million annually. The cost of new blanket oxide fuel
was based on the high range in Table 2.12 of Ref. 5. The high range,
$500/kg, was used because of the relatively low quantity to be pur-
chased. Hardware costs, estimated at $50,000 per fuel assembly, were

 

*Costs were not available from ANL on the IFR fuel cycle facilities
during the period in which this report was being prepared. However,
since that time ANL has developed a concept and cost estimates. ANL
should be contacted for needed details of this information.
E-14

based on experience at the FFTF and the EBR II. The operating costs
were scaled using a sizing exponent of 0.7.

The cost summary is given in Table E.2. This information suggests
that there is no significant economic advantage in a metal-fuel or an
oxide—-fuel FBR based on the costs of reprocessing and refabricating the
fuel.

Calculations were made to provide comparable values for processing
oxide fuel in a large fuel cycle facility (1500 t/a), assuming that it
existed and that the fuel was both reprocessed and fabricated there.
The cost basis was taken from Table 2.13 of Ref. 5. The results showed
costs of 4.7 to 6.3 mills/kWh for 80% reactor capacity factors.

It is obvious that the costs derived here for a 35-t/a facility are
high when converted to the unit cost ($/kg) basis or to mills/kWh, and
when compared to costs in a large facility. This is a function of the
low capacity requirements for either of the processes. The 25-35 t/a
plants are a factor of 50 smaller, and the cost values reflect this;
however, the cost comparisons should be more valid than the absolute
values, since both the metal and the oxide processes were costed on
approximately the same basis. Options other than integral reprocessing
have been suggested, as follows:8

l. Store the spent fuel until sufficient quantities are accumulated
for large-scale commercial reprocessing.

2. Ship the spent fuel to a large international reprocessing facility
set up as a cooperative venture.

3. Reprocess fuel in a small dedicated experimental facility such as
BRET, subsidized by research and development funds (for early power
plants).

4. Reprocess IMFBR fuel in a joint facility with LWR fuel (the
"hybrid” concept).

5. Reprocess IMFBR fuel in existing U.S. reprocessing facilities,
utilizing part of their capacity for c¢ivilian purposes, after
appropriate modifications are made.

E.4 ISSUES THAT NEED ATTENTION

The colocation of fuel reprocessing facilities with fast reactors
can require that all regulatory concerns for both the reactor and its
associated reprocessing facility be addressed before approval of the re-
actor is obtained. We have defined several issues in each of six areas
of the reactor-reprocessing-refabrication combination and have added a
comment on each issue.
E-15

Table E.2., Cost summary for reprocessing
and refabricating metal and oxide FBR fuel®

 

 

 

Fuel type Reprocessing costs ($109)
Capital Annual Annua} Total annual
capital operating
Metal fuel (20-year) 184—336 42—76 13 55—89
25.5 t/a (30~year) 184—336 3156 13 44—69
Oxide fuel (20-~year) 270—492 61—111 16 71127
35 t/a (30~year) 270492 45— 82 16 61— 98

Refabrication annual costs ($106)

 

Capital and

 

 

 

 

. Hardware Purchases Total
operating
Metal fuel (20-~year) 48--80 10 — 58—90
25.5 t/a (30~-year) 39—61 10 —_— 49—71
Oxide fuel (20~ or 25.6 - 13.5 39
8 t/a 30~year)
Combined annual costs ($106)
Repro. Refab. Waste Total
Metal fuel (20-year) 55—89 58—90 9.1 122—188
(30-year) 44—69 49—71 9.1 102—149
Oxide fuel (20-year) 77127 39 9.1 126176
(30-year) 61— 98 39 9.1 110—147
Busbar costs at 80% reactor capacity
(mills/kWh)
Metal fuel (20-year) 13.4—20.6
(25.5 t/a) (30~-year) 11.6—16.3
Oxide fuel (20~year) 13.8—19.3
(35 t/a) (30-year) 12, 116.1
Oxide fuel (20-Year) 6.3
1500 t/a  (30-year) 4o7

 

dA constant waste cost of 1 mill/kWh, or $9.1 million per year,

is assumed.
l.

E-16

Waste — One of the major questions that needs to be resolved for

the metal fuel processing concept is the means of handling the

waste. Before this question can be answered, considerable experi-
mental work is required to define where each of the major waste
isotopes will reside as a result of these processes. Once the
locations and the quantities of the wastes are defined, specific
disposal methods can be determined.

In the area of waste disposal, there are many questions relative to
the specific criteria for the wastes and waste containers for
either metal or oxide fuels. Such questions would require clar-
ification before final design could be completed. For the metal
fuel, considerable development work would be necessary to verify
proposed processes, particularly in the area of imperviousness of
the disposal product.

Environment — The environmental requirements for the total fuel

cycle need to be clarified before the experimental work is com-—

pleted, particularly for metal fuel, since little experience and no
source terms are available. This would require an early dedication
to, and funding for, such a project.

Again, clarification is needed as to what specific criteria or
regulations would apply to the releases from an FBR fuel cycle.
The current regulations (40CFR190) apply specifically to the
"uranium fuel cycle” for 1light-water reactors. It is not clear
whether 40CFR61 was intended to cover fuel reprocessing from non-
uranium fuel cycles (Par. IV B). This needs clarification.

Safety and Licensing — At present, there are no approved design

 

criteria for IMRs or for the fuel cycle facilities. Can these be
approved in time to be included in the design process?

At present there are also no regulations (NRC) which define the
criteria for fuel cycle facilities. Appendix P to 1OCFR50, which
has been proposed (but not approved), contains safety criteria
based on aqueous processing. It could be modified to cover pyro-
processing; however, this would have to be done well in advance of
the final design of a demonstration facility, in order for the re-
quirements to be included in the final design.

It is not clear that the safeguards requirements for the metal fuel
reprocessing and fabrication would be the same as those for similar
operations on the oxide fuel. However, the metal fuel would always
be highly radioactive and thus less attractive for diversion. For
either the oxide or metal fuel cycle, transportation is reduced by
using the integral fuel cycle.

Would any safety or licensing waivers be required for either FMEF
or FCF in order for them to meet DOE requirements, and could these
waivers be obtained?
E-17

Public Acceptance — Public acceptance has been a major problem for

 

less complex projects than these that are proposed. Could this be-
come a major problem?

Although the public may accept a reactor or several reactors at a
site, the addition of a fuel cycle facility to this "nuclear
park,” may require a great deal more education than before. This
could be particularly important for the management and disposal of
wastes.

Industry Acceptance — A major consideration will be the willing-

 

ness of industry to become involved in a complex, total-power pro-
gram such as that proposed for the IFR. How would industry propose
to handle this?

A. In the course of this study, we have observed that utilities
would not feel comfortable with staffing and managing the
diverse facilities required for a total fuel cycle. Their back-
ground and training are related to power generation and do not
include the complexities, costs, and risks of these other
elements. It may be more practical to have a separate operating
organization with the appropriate technical expertise.

B. Are the economics of omne central fuel cycle compatible with a
limited- (one-) reactor complex during early years when only one
or two reactors are in operation?

A fuel cycle facility must be originally designed and built to
handle the total output from all the reactors at the site.
Therefore, the major costs — those of construction — would be
incurred at some time before the total capacity of the facility
was needed. The introduction of the fuel cycle might be delayed
somewhat by storing the fuel after discharge. This would pro-
vide an inventory for the future reactors. An economic balance
would have to be made in order to define the optimum storage pe-
riod. The normal plant life for reprocessing is assumed to be
20 to 30 years, as opposed to an assumed reactor life of 30 to
40 years, however, the optimum lifetimes for both reactors and
reprocessing plants require further study and may prove to be
much longer.

C. How large an analytical complex will be required?

Reactors, per se, do not require an extensive analytical capa-
bility. The overall complex would require extensive remotely
operated facilities for determining chemical, metallurgical,
physical properties, and for analytical chemistry measurements.

D. Does the operator have sufficient backup technical capabilities
available (analytical laboratories, metallurgical caves, etc.)
to help resolve day-to-day problems?
E-18

The support facilities required to service a reactor fuel cycle
complex would be much more extensive than those required for a
reactor alone.

E. Can reactor(s) be dependent on only one fuel source and one dis-
posal method? How do reactors operate in case of a 1- to 2-year
shutdown of fuel cycle?

All of the facilities involved in the reactor cycle — fuel
source, fuel storage and disposal, and waste disposal — are
remotely operated and maintained. They would be difficult to
duplicate on short schedule in case a major problem developed in
any part of the system. Thus, the metal fuel reactors would not
have a fall-back option, unless and until several similar facil-
ities were available. The latter circumstance would require
shipment of spent and refabricated fuel.

F. What is the design life of the fuel cycle complex? Can it be
refurbished — and how often?

The design life of fuel <cycle facilities is normally 20-
30 years. The Savannah River Plant is 30 years old and still
operating well. There are no other plants that have operated
longer. It would appear feasible to replace equipment and ser-
vices within a plant to extend longevity, providing remote
handling capabilities were adequate.

G. How much plutonium is needed for the reactor program, at what
rate would it be supplied, and from what source?

Two questions are concerned with the availability of fuel (and
plutonium) for initial startup. First, is plutonium available
to provide enough metal fuel (e.g., one fuel assembly per week)
to support the pilot—-plant program for the 2 to 3 years that is
necessary to develop the process and waste-process system? And
second, what is the source of the plutonium and the manufactur-
ing capability to produce the initial two to three cores that
would be required for either the metal- or oxide-fuel demonstra-
tion reactor prior to recycling fuel from the fuel cycle facili-
ties? Will there be any problem in finding a vendor who will
gear up for such a limited program?

6. Reactor Interfaces

 

A. Can a reactor operate with the higher plutonium concentrations
that may be necessary to compensate for the less-fissile plu-
tonium due to recycle, or does the addition of the blanket fuel
compensate for this? (Would this affect the melting point of
the fuel?)
E-19

With each recycle of fuel, the nonfissile isotopes of plutonium
increase; thus, higher—quality plutonium may be required to pro-
vide the requisite fissile loading. A higher plutonium concen-
tration, without other adjustments, could result in a lower
melting point for the fuel. This would have to be evaluated
from the reactor safety viewpoint. These changes may be magni-
fied during the initial phases, when the fissile material is
changing from 235y to plutonium.

B. How large a variance in specifications of fuel will the system
accept?

Variations in metal fuel alloy composition and fissile composi-
tion can affect reactor performance. They can also affect the
precision of the burnup calculations, which may be the input
values for the fuel cycle SNM balance. Previous reactor cores
(albeit experimental in nature) have had high accuracy and pre-
cision requirements for fuel composition. In this case, both
the analytical determinations and the final adjustment will be
accomplished by remote means, which may not produce the designed
uniformity.

C. Is the fuel to be specified by percentage of plutonium, percen-
tage of fissile plutonium or total reactivity? How is this to
be determined?

If reactivity measurements are required, then special instrumen-
tation will have to be developed.

D. How are uranium/plutonium concentrations controlled to meet re-
quirements of (B) and (C) above?

A separate step may be needed between reprocessing and refabri-
cation which would permit composition adjustment.

E.5 Conclusions

The flowsheets described here appear to provide a viable process
for either the oxide- or the metal-fuel cycle. It is recognized that
additional development work is necessary for the metal fuel cycle, but
this does not appear to present a major problem. Although some areas
within the metal fuel reprocessing-waste handling flowsheets may change
as more development work 1is done, there is no reason to expect these
modifications to affect our overall conclusions.

It must be emphasized that the cost values presented here are gen-—
eric in nature and may be more accurate in a comparative sense than in
an absolute sense. It must also be pointed out that the principal cost
advantage of the wetal-fuel cycle (i.e., the relatively small number of
E-20

steps required for refabrication), is largely offset by blanket fabrica-
tion costs. The blanket for the metal fuel must be refabricated
remotely, while the oxide blanket can be fabricated by an existing LWR
fuel fabrication using contact means, which is considerably less expen-
sive. Because of these considerations, no appreciable difference was
found between the overall costs for the metal-fuel cycle and those for
the oxide—-fuel cycle. Although the calculated unit or bus bar costs for
a small (35 t/a) integrated fuel cycle do not appear to be attractive,
the corresponding fuel cycle costs for a projected large (1500 t/a)
facility appear to be competitive. A much more detailed cost study
would be required to refine the values presented here. Consideration
must also be given to the identified potential regulatory, social, and
industrial acceptance issues.

As discussed, the schedules presented here indicate that either the
metal~fuel or the oxide-fuel cycle could be closed within the allotted
time frame (2000-2010). Although the schedules contain leeway to pro-
vide for small delays in obtaining funding or in clarifying development
uncertainties, there are several technical and institutional problems
which must be resolved early for any integral reactor and fuel cycle
facility to meet the projected time frame.
E-21

REFERENCES FOR APPENDIX E

This flowsheet was developed from that initially provided in ANL-
IFR-7, Stream Flows and Compositions for Preliminary Pyroprocess
Flowsheets, March, 1985. The flowsheet includes modifications found

in the ANL-IFR series of reports and provided by ANL staff in

personal communications.

Preliminary Safety Analysis Report, Vol. 4, Nuclear Fuel Recovery

 

and Recycling Center, XN-FR-32, NRC Docket No. 50-564, Exxon Nuclear

 

Co., Inc. (1977).

M. J. Lineberry, R. D. Phipps and J. P. Burelbach, Commercial-side

 

IFR Fuel Cycle Facility: Conceptual Design and Cost Estimate,
ANL-IFR-25, October 1985.

D. H. Nyman, "Secure Automated Fuel Fabrication, Proceedings, 31st

 

 

Conference on Remote Systems Technology, 2, (1983).

Je. G« Delene, H. I. Bowers, and M. L. Myers, A Reference Data Base

 

for Nuclear and Coal-Fired Power Generation Cost Analysis, DOE/NE

 

0044R (1985).

Nada Stanie, Nucl. Fuel 13-14 (July 1, 1985).
W. D. Burch, private communication, July ]1985.

R. Balent and J. M. Yedidia, Draft, Large-Scale, Prototype Breeder

 

Fuel Cycle Plan, DOE/EPRI, July 1985.

 
APPENDIX F
860 MW(e) LARGE HIGH TEMPERATURE

GAS-COOLED REACTOR (HTGR)

The 2240 MW(t) reactor core is contained within a prestressed con-
crete reactor vessel (PCRV), with the core in the center cavity and the
steam generators and auxiliary heat exchangers in pods in the PCRV sur-
rounding the core.! The core is cooled with pressurized helium, moder-
ated and reflected with graphite, and fueled with a mixture of uranium
and thorium. It is constructed of prismatic hexagonal graphite blocks
with vertical holes for coolant channels, fuel rods, and control rods.
Helium coolant flows from four electric-motor-driven circulators down-
ward through the core, through four steam generators, and back to the
circulators. Superheated steam produced in once-through steam genera-
tors is expanded through a tandem compound turbine generator.

In addition to the four primary coolant loops, three core auxiliary
heat removal system loops are also provided. Each loop consists of a
gas/water heat exchanger with an electric-motor-driven circulator
located in a cavity in the PCRV wall. Should the main loops not be
available, coolant is circulated from the reactor core through the aux-
iliary heat exchangers where heat is transferred to the core auxiliary
cooling water system for eventual rejection from cooling towers to the
atmosphere.

The average core power density is about 6 kW/1 and the operating
pressure is about 7 MPa. The coolant gas exits the core at about
690°C. Steam conditions at the turbine inlet are 17.3 MPa and 541°C
providing a thermal efficiency of 387. The PCRV and ancillary systems
are housed inside a reactor containment building, which is a
conventional steel-lined reinforced containment structure. Typically,
balance-of-plant systems are housed in separate buildings depending on
function and service.

The advantageous safety characteristics of the large HTGR are based
on the high heat capacity of the graphite core and reflector, the high
temperature capability of the fuel and moderator, the use of a coolant
which does not change phase and has no reactivity effect associated with
density changes, the inherent shutdown mechanisms associated with a neg-
ative temperature coefficient, and the use of a PCRV which is a
redundant structure that precludes catastrophic failure. The low core
power density in combination with the graphite moderator leads to rela-
tively slow fuel temperature rises following loss—of-cooling accidents;
the graphite moderator and the ceramic fuel are stable to very high tem-
peratures, providing a high degree of fission product retention within
the fuel coatings up to about 1600-1800°C, and with only limited release
up to about 2000°C. The helium coolant does not undergo chemical reac-
tions within the reactor circuit, and the use of a gas coolant provides
unambiguous coolant conditions. Further, the large negative temperature
coefficient of reactivity for the fuel makes fast-acting shutdown
systems unnecessarye.
F-2

Nonetheless, if there is a complete loss of forced convection under
depressurized conditions, the afterheat generated in the core would
eventually cause plant damage and significant fuel particle coating
failures, since fuel temperatures would rise to values greater than
2000°C. As a result, engineered safety systems are used to supplement
the inherent characteristics of the reactor and include the independent
auxiliary cooling systems, independent and emergency reactivity shutdown
systems, and the reactor containment building.

REFERENCES FOR APPENDIX F

l. Gas-Cooled Reactor Associates, HTGR Steam Cycle/Cogeneration

 

2230 MW(t) Lead Plant, Conceptual Design Summary Report, HCR-20101,
February 1985. APPLIED TECHNOLOGY.
APPENDIX G

EVALUATION OF CLAIMS FOR THE MODULAR
HIGH TEMPERATURE REACTOR (HTR)

This appendix summarizes safety and economic claims which have been
examined by NPOVS for the modular HTR.

1. Modular HTR Safety Claims

 

The basic safety claims of the modular HTR are:

® doses to the public will not exceed values which would require
public evacuation [according to EPA's Protective Action Guide-
lines (PAG)] to a frequency of greater than 5 x 1077 per plant
year.

® Meet the NRC interim safety goals.

Further, to achieve these goals, a filtered confinement, rather than a
conventional containment is sufficient.

As specified under R&D needs, Section 3.5.3.1, additional source
term data will be needed to assess this claim.

Relative to investment protection, the claim is that the cumulative
frequency of events leading to plant loss is less than 107°/plant
year. This is below the NPOVS criterion No. 2.

2, Modular HTR Claims for Core Heatup Accidents (with Scram)

a. Circulating activity and fission product plateout are sufficiently
low and fuel performance during normal operation is sufficiently
good that doses can be maintained below EPA's guidelines for public
evacuation in core heatup accidents so long as the fuel is main-
tained below 1600°C.

To verify this claim requires additional data relative to the
source term as is specified in the R&D needs, Section 3.5.3.1 of
this report.

b. With loss of the main circulator or loss of cooling water flow to
the steam generator, and with loss of the auxiliary heat removal
system, heat transport to the vessel cooling system limits the fuel
temperatures to 1200°C. If the primary system is depressurized,
heat transport to the vessel cooling system limits fuel tempera-
tures to 1600°C. Further, under neither circumstance will there be
component damage.,
This claim is essentially supported by several analyses in the
United States and West Germany. Further sensitivity studies are
needed to examine the impact of uncertainties in various parameters
on temperatures of fuel and vessel internals.

Ce With a loss of all active and passive engineered cooling systems,
heat transfer to the earth limits the fuel to 1600°C. There would
be component damage (reactor vessel and internals).

This 1is dependent on soil properties and would require con-
firmatory analysis for each site.

3. Modular HTR Safety Claims for Transients without Scram

Although the core is provided with highly reliable primary and
reserve shutdown systems, rapid insertion of control material is not
required for core heatup transients involving the following:

® loss of cooling water flow to the steam generator,
° loss of forced circulation of helium,
o depressurization.

For these events which lead to core heatup, the increase in fuel tem-
peratures, combined with the negative temperature coefficient of reac-
tivity drives the reactor subcritical. Peak temperatures during the
core heatup are not significantly greater than with scram. After
several hours, depending on the specific thermal transient considered,
the fuel will have cooled sufficiently and the xenon will have decayed
sufficiently to cause recriticality.

Evidence that this claim can be met for conditions involving loss
of forced helium circulation is provided by tests which have been
performed at the AVR.!

4, Modular HTR Safety/Investment Protection Claims for Water
Ingress

Modular HTR proponents claim that water ingress is not a public
safety concern for the following reasons:

® the reactivity inserted can be compensated by the control and
shutdown systems

o the chemical reaction between water and graphite is a self-
limiting, endothermic reaction
G-3

® no flammable or combustible gas mixture will be produced in
the confinement since the primary system safety relief valve
will 1lift and discharge such gas through particulate filters
and up a stack into the atmosphere

Further, the claim is made that a long outage would be prevented either
by active engineered systems (moisture monitoring, steam generator iso-
lation and dump), or by a combination of these systems and operator
actions.

To examine the safety aspects of this claim will require computa-
tion of temperature coefficients and control rod worth under conditions
of water ingress, computation of the reactivity insertion and insertion
rate due to water ingress, and analysis of the resulting thermal
transient. An analysis of the chemical reaction and its consequences
should also be performed for the specific materials, geometry and
assumed sequences of events.

5. Modular HTR Safety Claims for Air Ingress

 

A serious air ingress accident is considered by proponents to be an
extremely unlikely event (K5 X 10~7 per reactor year) and thus is not
treated as a design basis event. The claim is that excessive oxidation
of graphite and resultant fission product release can occur only if
there are multiple ruptures in the primary system, there is no forced
cooling, and much more than one enclosure volume of air reacts with the
graphite. If there is forced cooling capability, even with a mixture of
helium and air at atmospheric pressure, core temperatures can be reduced
to 400°C in a few hours. The rate of oxidation at 400°C and below is
very low and no longer a safety concern. Finally, with a large air
ingress, the heat generation by oxidation is very small relative to
decay heat generation, so core temperatures are not expected to be sig-
nificantly greater than for a depressurized core heatup accident.

To examine the licensing aspects of an air ingress event, it may be
necessary to determine the risk associated with air ingress and to com-
pare that risk with the total risk from normal operation of the plant.
To examine the consequences of air ingress requires analytical model
development (currently in progress) for accident simulation, an under-
standing of the gas exchange between the primary system and the confine-
ment and a thorough understanding of the oxidation process. Further
provisions in the confinement to limit air accessibility should be
examined for practicality.

6. Economics and Constructibility Claims

 

a. Claim. Total power generation costs are competitive with coal
plants of equal capacity.
G-t

Evaluation. Because of the preliminary conceptual nature of the
design, there are large uncertainties, especially in capital investment
costs. Certainly, it can be claimed that this concept holds significant
promise to be competitive. Design and cost studies should continue to
better define the economic competitiveness with coal.

It is recommended that studies be carried out to provide improved
capital cost estimates by:

(1) Continued development of the design to the point that quantities of
commodities and labor can be estimated and compared with current
LWR and coal-fired plant experience.

(2) Development of estimates of indirect costs (e.g., manhours of
design engineering and project management, instead of relying on
percentages).

(3) Starting with a "first-of-a-kind” plant, development of the
strategy for arriving at the cost of an "Nth-of-a-kind plant.

b. Claim. Significant capital investment cost savings are
achieved through use of a confinement, rather than containment.

Evaluation. Bechtel made a study in FY 1982 (Ref. 4) in which the
added direct cost for containment for a modular HTR plant with 8 reac-
tors was estimated to be $§70 X 106. Escalation to 1985 and addition of
indirect costs increases this amount to about $120 x 106 or $150/ kW(e),
which is a significant cost.

c. Claim. Systems outside the nuclear island can be procured and
installed to non-nuclear standards resulting in an overall savings in
capital investment cost of approximately 10% (Ref. 5).

Evaluation. It has been documented®»7 that the cost of non-nuclear
portions of LWR plants is much higher than for coal-fired plants. The
reasons cited are as follows:

® Nuclear quality standards affect the attitudes of all persons
working on the entire project — management, engineering, and
crafts.

® Bulk materials for the entire project, such as rebar, anchor

bolts, embedments, small bore piping, and concrete, are pro-
cured and handled as required by a nuclear quality assurance
program to eliminate danger of degrading the quality of
safety-related structures and systems by inadvertent sub-
stitution.

° Non-safety structures adjoining safety-related structures are
designed to prevent collapse in the event of a design Dbasis
earthquake or tornado.
o Management and supervision are preoccupied with problems asso—
ciated with safety-related facilities and often neglect
planning for non-safety facilities.

Bechtel addresses these problems by proposing to provide physical sepa-
ration between the nuclear and non—-nuclear facilities and between the
construction forces so that the low productivity experienced in nuclear
construction is not transferred to the non-nuclear areas. Our analysis
as applied to current IWRs confirms the 10% level of savings in invest-
ment cost.® However, there are also reasons why these savings may not
be fully realized:

° Dispersion of plant facilities with longer ruas for piping and
wiring and cables.

® Duplication of construction management and construction
facilities.

The entire question of separation of facilities to achieve increased
labor productivity is also a management issue., For example, a few U.S.
utilities have been able to build 1WR plants with about one-half the
labor content of the average LWR plant.’ R&D should be directed toward
determining how bhest to manage and organize nuclear construction pro-
jects to assure higher labor productivity. Separation of construction
is one organizational approach to promoting better management.

d. Claim. Availability can exceed 807% with futher design improve-
ments. Use of two or four turbine generator sets could increase avail-
ability.?

Evaluation. HTR Program availability studies are continuing in the
DOE program. Achievement of such values depends upon the unscheduled
outages which will occur, as well as on refueling and scheduled outage
times. Estimates of unscheduled outages and refueling times are
uncertain in our evaluation; while >»80% availability appears to be
achievable (see below), it cannot be assured at this time.

The North American Electric Reliability Council (NERC) Equipment
Availability Report for the l0-year period 1974—1983 supports the claim
of smaller turbine generators contributing to higher overall plant
availability. As shown in Table 3.4 of Volume III, fossil plant
turbine—-generator sets in 400 MW(e) and below sizes have distinctly
lower forced and scheduled outage rates and higher availabilities than
turbine—generator sets in the larger sizes. The advantage is even more
significant for nuclear plant turbine-generator sets below 800-MW(e)
size, compared with those above 800 MW(e). It also is observed that
nuclear plant turbine—-generator sets have a distinct performance
advantage over fossil sets in all size ranges., It is speculated that
this may be due to the lower steam temperatures and pressures and
rotational speeds of nuclear turbine-generator sets, which result in a
less severe operating enviroament.
G-6

As shown in Table 3.5 of Volume III the area having the greatest
potential for LWR plant availability improvement is with the reactor
and associated systems. It is obvious that plant availability improve-
ment R&D must concentrate on the reactor and its related systems. For
example, a 50% reduction in reactor scheduled outage factor plus a 50%
reduction in reactor forced outage rate would result in 80% overall
plant equivalent availability. The above LWR data provides strong
support that the modular HTR goal can be achieved.

e. Claime. A four-module plant can be constructed in 36 months
from start of site work to commercial operation of the second turbine.?

Evaluation. This is an optimistic schedule when judged by U.S.
experience, but appears possible. The modular HTR schedule is based on
an evaluation of the conceptual design by Bechtel. It should be re-
examined after the design has been carried to point of estimating quan-
tities of construction materials (structural steel, concrete, piping,
and wiring,) and labor manhours. This will provide a firmer basis for
estimating elapsed time for placing equipment and materials. The
construction schedule should be reevaluated as the design progresses.

f« Claim. The plant can be operated and maintained by a staff of
306 (Ref. 2).

Evaluation. This estimate is based on a preliminary analysis of
staffing requirements. In comparison, typical staffing for current
large IWR plants is ~400. The modular HTR staffing estimate is based on
the following assumptions:

o Regulatory procedures have been stabilized (i.e. no back-
fitting by maintenance forces).

° Plant control is highly automated, permitting operatiom of
four reactors by one control operator station.

° Plant is designed for maintenance with one module offline re-~
sulting in minimum requirements for peak maintenance forces.

° Plant security is highly automated, requiring minimum security
forces.

These assumptions, along with the smaller turbine-generator size requir-
ing less maintenance personnel, are the principal reasons for the re-
duced staffing. Further studies of operation and maintenance staffing
should be performed, using a task—analysis approach, especially in the
area of control operation of multiple reactors by a single control oper-
ation and its relation to safety.

e Claim. Adding capacity in small increments 1is/will be a
significant financial goal of utilities and results in less financial
risk.
Evaluation. In a time period of low load growth, high financing
and construction costs, and reluctance of public utility commissions to
grant rate adjustments, this claim appears intuitively obvious. Two
recent draft studies by Los Alamos National Laboratory (LANL) and
Applied Decision Analysisg’9 support this claim. Both studies attempted
to quantify the additional capital investment cost that utilities could
afford to pay for smaller, shorter lead time plants in comparison with
larger, longer lead time plants, while continuing to meet their
financial goals. LANL found that a reduction from long to medium lead
times permits the utility company to pay 40—50% more in overnight
construction costs, and a reduction from 1long to short lead times
permits a four-fold increase in the overnight construction cost. From a
ratepayer viewpoint, Boyd et al.? found that utilities could pay approx-
imately $200/kW(e) capital investment cost premium for smaller unit
sizes and shorter lead times for utility system sizes 3000 MW(e) and
larger, From the shareholder viewpoint, the affordable capital invest-
ment cost premium was found to be two to three times higher than from
the ratepayer viewpoint. The findings are general in that they apply to
any type of power plant, they are also sensitive to a number of
parameters f{(e.g., system size, existing generation mix, load growth
rate, and financing). However, sensitivity analyses in both studies
support the claim. We recommend, however, that the studies in this area
be continued and refined to develop a complete understanding of the
economics of small nuclear plants.
G-8

REFERENCES FOR APPENDIX G

1,

2.

3.

7.

8.

K. J. Kruger and G. P. Ivens, “"Safety-related Experiences with the
AVR-Reactor,” presented at the IAEA Specialists' Meeting on Safety
and Accident Analysis for Gas-Cooled Reactors, 0Oak Ridge, Tennes-
see, May 13-15, 1985.

Preliminary Concept Description Report, 4 x 250 MW(t) HTGR Plant
Side-by-Side Steel Vessel Prismatic Core Concept, HIGR-85-142,

 

issued by Bechtel Group, Inc., et al., for Gas-~Cooled Reactor
Associates, San Diego, California, October 1985, APPLIED
TECHNOLOGY.

HTGR Program Concept Evaluation Plan for Small HTGRs, GCRA 84-009,

 

Gas-Cooled Reactor Associates, San Diego, California, October 31,
1984,

Modular HTGR Balance of Plant Design and Cost Status Report,

 

Bechtel Group, Inc., San Francisco, California, September 1982.

Modular HTGR System Design and Cost Summary, GCFR-00693, Bechtel
Group, Inc., San Francisco, California, September 1983,

Constructibility Assessment for Modular High-Temperature Gas-Cooled

 

Reactors, Bechtel Group, Inc., San Francisco, California, July

1982.

Phase IV Update of the EEDB, DOE/NE-0051/1, United Engineers and

 

Constructors, Inc., Philadelphia, Pennsylvania, September 1984,

Andrew Ford, The Market for New Electric Generating Capacity: A
Financial Feasibility Case Study, Los Alamos National Laboratory,

 

 

Los Alamos, New Mexico, June 1984,

D. W. Boyd et al., The Potential Impact of Modularity vs. Utility
Generation Investment Decisions, Decision Focus, Inc. and Applied

 

 

Decision Analysis, Inc., for the Electric Power Research Institute,
Palo Alto, California, March 1984.
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15-16.
17.
18.
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20-24,
25,
26,

101.
102-177.

178-283.

INTERNAL DISTRIBUTION

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Dodds, Jr.

27,
28,
29.
30.
31-47,
48,
49,
50.
51.
52,
53.
54-73,
74.
75.
76=-93,
9.
95.
96.
97.
98-99.
100.

ORNL/TM-9780/V2

E. Jones Jr.
R. Kasten

C., Maienschein
P. Malinauskas
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. Moses

. Mynatt

. Oakes
Rayner

Selby

. Trammell

. Trauger

Uhrig

. Wehe

. White

. Wilbanks

Wymer

Central Research Library
Document Reference Section
Laboratory Records Department
Laboratory Records (RC)

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EXTERNAL DISTRIBUTION

Office of Assistant Manager for Energy Research and Development,
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Distribution Category UC-79T, Liquid Metal Fast Breeder Reactors:

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Nuclear Power Options Viability Study Distribution.

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