^'pTDa|Or/'/ &*¦»?
PB-209 283
Survey of Nuclear
Power Supply Prospects
Hittman Associates, Inc
February 1972
National Technical Information
U. S. DEPARTMENT OF COMMERCE
5285 Port Royal Road, Springfield Va. 22151
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BIBLIOGRAPHIC DATA 1. lUportNo. 2.
SHEET APTD-1077
3. Ktcipicnt's Acclv urn \o.
PB 209-283 ,
4. I illi and Subtitle
Survey of Nuclear Power Supply Prospects
5. Hcpori Date
February 1972
6.
7. Author(s)
8. Performing Orgaiii/Mtmn Iti p>
No- HIT-501
9. Performing Organization Name and Address
HITTMAN ASSOCIATES, INC.
COLUMBIA, MARYLAND 21045
10. Pro|cct/l ask/Work Unit No.
11. Contract/Grant No.
EHSD-71-43
12. Sponsoring Organization Name and Address
ENVIRONMENTAL PROTECTION AGENCY
Office of Air Programs
13. Type of Report & Period
Coveted
14.
IS. Supplementary Notes
16. Abstracts A detailed review of the nuclear segment of the power industry and an evaluatio
of the ways in which nuclear power will impact on national air quality and emissions from
fossil-fueled steam electric plants are presented. Electrical power from nuclear-fueled
reactors will continue to grow at a substantial rate and will ultimately supply a major
portion of electrical power in the United States. The development and use of nuclear plan
will di.rectly affect the use of fossil fuels to supply electrical power and will also di-
rectly affect the amounts of combustion products released to the environment. Forecasts
of the proportion of nuclear power, and thus of fossil-fueled power, vary widely although
the maj&rity of forecasts are in reasonable agreement through about 2020. The forecast of
future power supplied by nuclear reactor^ has been shown by optimization analysis to be
strongty•dependent upon the reactor types and their technological development. Depending
on the placement in service of various reactor systems, the emissions of sulfur dioxide
and other pollutants to the atmosphere may vary by a factor of 10 within the next 80 year
independent of the allowable limits of emissions orfuel contents .,Jfoss i 1-fueled generat-
ing plants will continue to produce power at levels'equal to or greater than 1970 levels
-for the next 80 years, therefore,"emission* reductions-can~orvly be-ach hevet) through contro
devices-or fuel modif Ic^t-fons to remove undesirable emissions from effluent gases. Light
water reactors will provide the bulk of nuclear-fueled power through the next 15 years,
although some use of advanced converter nuclear systems is projected, notably in the High
Temperature Gas-Cooled Reactor System. Breeder reactor technology will develop sufficient
ly in the next decade to enable utilities to place breeder reactors in commercial service
by the middle of the next decade. Fusion reactor systems are currently under development
but are not expected to become commercially feasible for the production of electrical pow
er before the first decade of the next century.
17. Key Words and Document Analysis. 17a. Descriptors
Air pollution Combustion products
Nuclear power plants Sulfur dioxide
Nuclear reactors
Fuels
17b. Identifters/Open-Ended Terms
17c. COSATI Field/Group 1 3B
Reproduced by
NATIONAL TECHNICAL
INFORMATION SERVICE
U S Department of Commerce
Springfield VA 23131
IB. Availability Statement
Unlimi ted
FORM NTl*aB 110*70)
19. Security Class (Ibis 21
Report)
iiUM.ASsiFirn
20. Security (..lass (Ihis
Pane
UNC1.ASSI
22. Pmc
UT
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SURVEY OF
NUCLEAR POWER SUPPLY
PROSPECTS
HIT-501
February 1972
Prepared Under
Contract No. EHSD 71-43
Environmental Protection Agency
Office of Air Programs
HITTMAN ASSOCIATES, INC.
COLUMBIA, MARYLAND
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u
LEGAL NOTICE
This report was prepared as an account of Government sponsored work.
Neither the United States, nor the Environmental Protection Agency, Office of
Air Programs (EPA-OAP), nor any person acting on behalf of EPA-OAP:
A. Makes any warranty or representation, expressed or implied with
respect to the accuracy, completeness, or usefulness of the information con-
tained in this report, or that the use of any information, apparatus, method,
or process disclosed in this report may not infringe privately owned rights;
or
B. Assumes &ny liabilities with respect to the use of, or for damages
resulting from the use of any information, apparatus, method, or process
disclosed in this report.
As used in the above, "person acting on behalf of EPA-OAP" includes
any employee or contractor of EPA-OAP, or employee of such contractor, to
the extent that such employee or contractor of EPA-OAP, or employee of such
contractor prepares, disseminates, or provides access to any information
pursuant to hia employment or contract with EPA-OAP, or his employment
with such contractor.
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iii
TABLE OF CONTENTS
Page
LEGAL NOTICE ii
TABLE OF CONTENTS in
LIST OF FIGURES iv
LIST OF TABLES v
I. INTRODUCTION 1-1
II. SUMMARY AND CONCLUSIONS II-1
III. BACKGROUND AND DEFINITION Ill-1
A. Fission Power Ill-1
B. Fusion Reactor Power 111-24
IV. EXISTING AND PLANNED GENERATING STATUS BY
TYPE AND REGION IV- 1
V. LONG-TERM PROSPECTS V-l
A. Comparison of Nuclear Power Projections V-l
B. Effects of Reactor Types and Their Development on
Energy Projections and Costs V-8
C. Fuel Availability V-l9
VI. SAFETY CONSIDERATIONS, SPECIAL LICENSING
REQUIREMENTS, DELAYS VI-1
VII. ADVANTAGES AND DISADVANTAGES OF NUCLEAR
POWER VII-1
A. Light Water Reactors VII-1
B. Advanced Converters VII-2
C. Thorium Systems VII-3
D. Breeder Reactors - . - . . VII-4
VIII. EFFECT OF NUCLEAR POWER ON NATIONAL AIR QUALITY. VIII-1
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iv
LIST OF FIGURES
Figure
No. Title Page
II-1 Percent Fuel Distribution 11-3
II-2 Actual Fuel Distribution II-4
III-1 Fusion Schematic III- 26
IV-1 Nuclear Power Reactors in the United States IV-11
V-l Growth of On-Line Nuclear Capacity V-4
V-2 Forecasts of U. S. Nuclear Power Requirements V-5
V-3 Forecasts of U.S. Total Electricity Requirements V-7
V-4 Cumulative Nuclear Generating Capacity V-14
V-5 Nuclear Percentage of Future Total Power Capacity V-l5
Installed After 1970
VIII-1 Potential Sulfur Dioxide Emissions from Fossil Steam VIII-5
Electric Plants
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V
LIST OF TABLES
Table
No. Title Page
II-1 Projected Cumulative Power Capacity by Fuel II-1
III-1 Types of Nuclear Power Reactors III-2
III-2 Summary of Design and Performance Parameters III-3
for Light Water Reactors
III-3 Engineering Characteristics of Advanced Converters III-9
III-4 Summary of Design and Performance Parameters for III-15
Molten Salt Reactors
III-5 Performance Characteristics of Liquid-Metal-Cooled III-18
Reactors
III-6 III- 22
IV-1 Central Station Nuclear Plants - Census Region: IV-2
New England
IV-2 Central Station Nuclear Plants - Census Region; IV-3
Middle Atlantic
IV-3 Central Station Nuclear Plants - Census Region: IV-4
South Atlantic
IV-4 Central Station Nuclear Plants - Census Region: IV-5
East North Central
IV-5 Central Station Nuclear Plants - Census Region: IV-6
East South Central
IV-6 Central Station Nuclear Plants - Census Region: IV-7
West North Central
IV-7 Central Station Nuclear Plants - Census Region: IV-8
West South Central
IV-8 Central Station Nuclear Plants - Census Region: IV-9
Mountain
IV-9 Central Station Nuclear Plants - Census Region: IV-10
Pacific
V-l Nuclear Forecast Functional Relationships V-3
V-2 Results of Systems Analysis Calculations for Fossil V-ll
and Nuclear Plants with Rising Uranium Costs
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vi
LIST OF TABLES (Continued)
Table
No. Title Page
V-3 Estimated U. S. Uranium Resources V-20
V-4 Quantities of U^Og Fuel RequiredEach Year V-21
V-5 Estimated U. S. Thorium Resources V-21
VIII-1 Capacity Forecasts by Fuel VIII-3
VIII-2 Tons of SO2 Emitted per Year VIII-3
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1-1
I. INTRODUCTION
Following the successful controlled fission reaction by Enrico Fermi
and his co-workers at the University of Chicago on December 2, 1942, the
immediate applications of nuclear power were devoted to the existing war
effort. Following World War II, scientific attention was directed toward the
application of nuclear energy to electrical power generation. This develop-
ment program led to the installation of the first commercially-operated
nuclear plant at Shippingport, Pennsylvania. This 90 megawatt plant, placed
on line in 1957, is owned by the Atomic Energy Commission and operated by
the Duquesne Light Company. Since that time, many more and larger nuclear
plants have been placed in service and are successfully producing ever-
increasing percentages of national power demands. At present, over 110
nuclear fueled electric generating units are in operation, under construction
or ordered.
Most of the nuclear plants presently planned for operation are of the
light water cooled and moderated variety. Long-term operating experience
with such plants as Shippingport (1957; 90 megawatts), Dresden 1 (1959;
200 megawatts), and Yankee Rowe (1962; 175 megawatts) has adequately
demonstrated the reliability and practicality of light water reactor systems.
In 1967, a high temperature gas-cooled reactor at Peach Bottom, Pennsylvania
was placed in commercial operation, and has demonstrated the practicality
of this concept.
Overall, nuclear power reactors are presently delivering almost three
percent of national electrical supply, nearly 10,000 megawatts from the
modest beginnings of 90 megawatts in 1957. Nuclear power has made
impressive inroads and contributions to the problems of supplying reliable
and pollution-free electrical power in the United States.
Nuclear energy systems are experiencing steady growth in sophistication
and in technological advancements and improvements. Current on-going
efforts are being directed toward developments in all aspects of nuclear power
systems, including future power requirements, economics, resource
conservation, reactor technology, safety, and long-range planning. As a
result of the Atomic Energy Acts of 1946 and 1954, the Atomic Energy
Commission has the major responsibility for regulating and promoting the
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1-2
use and development of nuclear energy to meet the expanding needs for safe
and reliable electric power. In cooperation with the utilities and other in-
volved organizations in the public and private sectors, the Atomic Energy
Commission is presently committed to the development of new reactor
technology, with primary priority focussed on the Liquid Metal Cooled Fast
Breeder Reactor (LMFBR). Simultaneously ongoing pro]ects include the
advanced convertor reactors such as the High Temperature Gas-Cooled
Reactor (HTGR) and the Fusion Reactor.
Where foreign research efforts, notably these of Canada in the areas
of heavy-water-moderated organic cooled reactors and BLW reactors, have
coincided with those of the United States or have been performed as part of
a joint technical effort with the United States, these have been described
briefly in this report.
The advanced convertor reactor technology has advanced to a point
where a demonstration plant (Peach Bottom, 40 Mwe} has been operating
for several years, and a commercial-sized plant (Fort St. Vrain, 330 Mwe)
will be placed on line in L972. From all indications, breeder reactors will
be built and ready for commercial operation about 1985. Potentially, the
breeder reactors, which will produce more fissionable material than is
consumed and which will operate at thermal efficiencies equal to or greater
than comparable fossil plants, will provide the utility industry with an
economically and technologically attractive means of producing electric
power. Breeders will in all likelihoodbecome a primary source of steam
for power generation, outstripping present fission reactors and fossil plants
in new plant construction once the breeder is sufficiently developed for
commercial application. Theoretically, a fusion reactor, which will utilize
the principle of fusing light elements into heavier elements with a release of
energy, is a possibility for future power production requirements. However,
the necessary technology has not yet been developed. It is unlikely that
fusion power reactors will be commercially available until after the year
2000.
The report following was prepared for the Environmental Protection
Agency, Office of Air Programs for the purpose of forecasting the nation's
future nuclear power supply prospects. Nuclear power is expected to gain an
increasingly larger share of the future power supply market because of its
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1-3
more attractive economic conservation and pollution picture as compared to
fossil fuel power. Sections I-IV of the following report detail the reactor
types now in operation, as well as those under development, and indicate
the existing and planned nuclear generating status of each state through 1980.
The final sections of this report evaluate several previous nuclear power
supply forecasts and formulate a forecast based on this previous work as
well as data collected for this report. The reactor types described in
Sections 1-1V are evaluated in terms of energy projections and cost, fuel
availability, safety considerations and advantages and disadvantages. The
beneficial effect of nuclear power on air quality is also detailed.
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II-1
II. SUMMARY AND CONCLUSIONS
For a variety of reasons, among them rising fossil fuel costs, increased
costs of air and water pollution control measures, and economies of scale, the
utility industry has been turning to nuclear powered steam plants to provide
base loading power supplies, particularly when larger sized plants are involved.
Of all new plants ordered for construction over the next several years, approxi-
mately 40 percent are nuclear fueled (Ref. II-1). It is expected that nuclear
plants will supply 19 percent of electrical power in 1980, 35 percent in 1990,
and over 45 percent in 2000.
At present, the primary reactor sales for nuclear power generation are
the light water types, Including the boiling water reactor and the pressurized
water reactor. In addition, high temperature gas-cooled reactors have also
been purchased for commercial operations. These reactor systems will pro-
vide basic nuclear steam supplies to the utility industry through about 1985,
when it is anticipated that breeder reactors will become commercially avail-
able for power generation. Thereafter, continuing developmental work will
ultimately provide the necessary technology for commercial application of a
fusion reactor system, although it is unlikely that a working fusion reactor
steam supply can be in service until sometime after the turn of the century.
From various forecasts of the quantity and distribution of steam supply
systems for electrical power generation, a single forecast of the utility industry
has been formulated (Ref. II-2) extending through the next 80 years. Based on
this information, a prediction of electrical generating capacity through 2050
by fuel types has been derived (Ref. II-1) and is shown in Table II-1.
TABLE II-1. PROJECTED CUMULATIVE
POWER CAPACITY BY FUEL (Mwe x 103)
Fuel
1970
1980
1990
2000
2010
2020
2030
2040
2050
Coal
176
280
412
566
640
615
595
510
416
Oil
22
34
49
60
76
58
52
45
40
Gas
61
84
108
119
115
103
93
85
80
Hydro
51
62
69
89
90
95
100
102
104
Nuclear
10
106
343
671
1080
1825
2660
3458
4560
Total 320 570 980 1500 2000 2700 3500 4200
5200
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II-2
It may be noted from Table II-1 that nuclear power plants are expected
to assume the major share of power production through the forecast period.
Nevertheless, fossil fuels will continue to grow in usage through about 2010,
when they are expected to commence to decline. Fossil fuels will, over the
next 30 to 40 years, provide a major portion of power supply but will decrease
in percentage from over 80 percent in 1970 to 70 percent in 1980, 58 percent
in 1990, and 50 percent in 2000. After 2000, total fossil fuel usage for elec-
trical power generation will decline on both percentage and actual bases.
Figures II-1 and II-2 illustrate these forecasted changes. Figure II-2 par-
ticularly illustrates the continued expected growth of the use of fossil fuels in
generating electrical power and indicates that a four-fold increase in fossil
fuel combustion will cause a continuing rise in combustion products to the
environment through about 2010, when a gradual decline is forecast to begin.
Because of the forecasted growth in the use of fossil fuels, and despite the
overwhelming forecasted growth of nuclear plants, there is no time within
the foreseeable future when fossil fuel use for power generation will be any
less than it is today and, in fact, after a 40-year decline beginning in 2010,
will only reach 1985 levels. Thus, it may be expected that, even with the
very large trend to nuclear fuels, fossil plants either in existence or yet to
be built will contribute increasing amounts of combustion waste products to
the environment through the middle of the next century at levels equal to or
greater than at present, barring any major technological advances to prevent
this occurrence.
^
This report includes a detailed review of the nuclear segment of the
power industry and an evaluation of the ways in which nuclear power will
impact on national air quality and emissions from fossil-fueled steam electric
sue (^ •—v
plants^ The followjjig' is a list of primary conclusions drawn from this report:
• ^—Electrical power from nuclear-fueled reactors will continue
to grow at a substantial rate and will ultimately supply a major
portion of electrical power in the United States.
• " The development and use of nuclear plants will directly affect
the use of fossil fuels to supply electrical power and will also
directly affect the amounts of combustion products released to
the environment.
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II-3
Figure II-1. Percent Fuel Distribution
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II- 4
Year
Figure II-2. ActuaJ Fuel Distribution
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II-5
Forecasts of the proportion of nuclear power, and thus of
fossil-fueled power, vary widely although the majority of
forecasts are in reasonable agreement through about 2020.
The forecast of future power supplied by nuclear reactors
has been shown by optimization analysis to be strongly depen-
dent upon the reactor types and their technological develop-
ment. ^Thus, several differing nuclear power projections
result from differences in the expectations for timely deploy-
ment of light water reactors, advanced converters, thorium
systems, breeder reactors, and combinations of these with
fossil-fueled plants.
^Depending on the placement in service of various reactor
systems, the emissions of sulfur dioxide and other pollutants
to the atmosphere may vary by a factor of 10 within the next.
80 years, independent of the allowable limits of emissions or
fuel contents. ^ This is directly proportional to the percentage of
total electrical power produced by nuclear and fossil steam
plants.
Fossil-fueled generating plants will continue to produce power
at levels equal to or greater than 1970 levels for the next 80
years. Therefore, emission reductions can only be achieved
through control devices or fuel modifications to remove
undesirable emissions from effluent gases.
Light water reactors will provide the bulk of nuclear-fueled
power through the next 15 years, although some use of advanced
converter nuclear systems is projected, notably in the High-
Temperature Gas-Cooled Reactor System.
Breeder reactor technology will develop sufficiently in the
next decade to enable utilities to place breeder reactors in
commercial service by the middle of the next decade. Breeder
reactors will comprise the largest majority of nuclear systems
constructed from 1985 through the beginning of the next century.
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Fusion reactor systems are currently under development
but are not expected to become commercially feasible for
the production of electrical power before the first decade
of the next century.
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II-7
REFERENCES
II-1. A Review and Comparison of Selected United States Energy
Forecasts, Pacific Northwest Laboratories, Battelle Memorial
Institute, Columbus, Ohio, December 1969.
II-2. Bibliography and Digest of U.S. Electric and Total Energy
Forecasts. 1970-2050, Publication 69-23, Edison Electric
Institute, New York, 1969.
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Ill-1
IE. BACKGROUND AND DEFINITION
A. Fission Power
The power reactors discussed in this section are classified into four
categories:
(1) Light water reactors
(2) Advanced converters
(3) Thorium systems, and
(4) Breeder reactors
In total, there are approximately 10 of these reactor types as shown in
Table III -1 - This classification follows that of the Systems Analyses Task
Force organized by the AEC Division of Reactor Development and Technology
to evaluate the various reactor types. These basic types may be further
subdivided as seen in the text to follow.
1. Light Water Reactors
This category of reactors includes the boiling water reactor (BWR)
and the pressurized water reactor (PWR). As can be seen from Table III-2
(Ref. III-l), the performance characteristics for the BWR and PWR are
similar. The basic source of these performance data is Reference III-2.
Boiling water reactors are currently manufactured by the General
Electric Company. They are sold as steam supply systems in a range of
sizes from 500 to 1100 Mwe.
The first large commercial application of the BWR concept was the
Dresden I reactor located near Morris, Illinois. This reactor plant was
initially rated at 180 Mwe and is currently operating at 200 Mwe. It
employed the dual cycle concept with an elevated steam drum and a recir-
culation system contained in a spherical steel shell containment structure.
Through March 1968 the plant was operated at an average availability factor
of approximately 90 percent. A recent BWR reactor design (1075 Mwe
net capacity) has the performance parameters shown in Table III-2.
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Ill-2
TABLE III -1. TYPES OF NUCLEAR POWER REACTORS
Classification
Reactor Type
Assigned
Acronym
1. Light water reactor
(LWR)
Boiling water reactor
Pressurized water reactor
BWR
PWR
2. Advanced converters
Heavy water-moderated HWOCR
organic-cooled reactor
Heavy water-moderated HWBLW
boiling-light water-cooled
reactor
High temperature gas-cooled HTGR
reactor
3. Thorium systems
Molten salt reactors:
Molten salt converter
Molten salt breeder
MSR
MSCR
MSBR
4. Breeder reactors
Liquid-metal-cooled fast
breeder reactor
Steam-cooled fast breeder
reactor
LMFBR
SCFR
Gas-cooled fast reactor
GCFR
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III-3
Table III-2 SUMMARY OF DESIGN AND PERFORMANCE PARAMETERS
FOR LIGHT WATER REACTORS
Reference
Reference
1980-1990
1990-on
BWR
PWR
LWR
LWR
Core Region:
Active height, ft.
12.0
12.0
Diameter, ft.
15.6
11.1
Nunber of control rods
185
53
Number of fuel elements
76U
193
Fuel pin diameter, inches
.562
.1»22
Number of fuel pins per element
I49
20U
Power, MWt
3293
3083
Average specific power, MWt/MT
22.0
3b. 8
39.9
ltU.8
Peak linear rod power, kW/ft
18.3
18.0
Fuel average discharge exposure, MWD/MT
27,500
30,000
30,000*
30,000*
Hot-spot cladding temperature, °F
565
657
Maximum fuel temperature, °F
U38O
I4100
Maximum heat flux, B/hr-ft^
U25,000
553,700
Minimum critical heat flux ratio
1.90
1.86
Core subcooling, Btu/lb
25.5
ll»7.8
Reactor pressure, Psi
1050
2235
Core inlet/outlet temperature, °F
376.i/5'*6.u
51*5/610
Average core power density, Kv/1
50.8
93.1
Average core exit quality, w/o
13.6
ri/U atom ratio, average, full power
3.71*
U.23
Specific inventory, Kg fissile/MWe
3.0
2.2
1.72
l.'il
Pu Yield, Kg fissile/MWe-yr, at avg.
exposure
.22
.23
.23
.23
U^Og consumption, Kg/MWe-yr
188
200
200
200
Thermal to electrical conversion
efficiency, %
32.8
32.5
32.5
32.5
Reference: WASH 1082, "Current Status and Future Tecnnieal and Economic Potential
of Light Water Reactcft-s," March 1968.
•30,000 MWD/MT data shovn here for comparison of performance vith reference designs
tut actual exposures used in calculations varied.
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Ill-4
Improvements were made in the design parameters of boiling water reactors
increasing the operating efficiency and reducing "the cost of generated power.
(a) More power delivered per unit weight of loaded uranium,
or greater specific power
(b) More power delivered per unit core volume, or greater
power density
(c) Improved exit steam quality
(d) Increased exposure of fuel in terms of integrated thermal
energy delivered per unit weight of fuel (MWD/MT)
These improvements result principally from improved optimization of
core power distributions, design innovations, improvements in anticipated
heat transfer characteristics and application of internal steam separation.
The BWR steam supply system includes the reactor vessel and its
internal components and all of the primary and auxiliary circulating equip-
ment such that the system can provide input steam to the turbine-generator
for producing power. Components within the reactor vessel incLude the core,
control rod drive systems, and jet pumps. The core consists of the fuel
assemblies, channels. controL blades, and instrumentation to monitor its
status at various stages of operation. Each fuel assembly consists of a
1x7 array of fuel rods enclosed in a Zircaloy channel. The fuel rods
consist of slightly enriched uranium dioxide pellets contained in a Zircaloy
tube cladding. The cruciform-shaped control rods occupy alternate locations
between fuel assemblies and are withdrawn into the guide tubes of the core
during operation in accordance with a predetermined pattern.
Pressurized water reactors are currently manufactured by the
Westinghouse Electric Company, Babcock & Wilcox Company and Combustion
Engineering, Inc. The design approach employed by these companies is
quite similar. The Babcock & Wilcox Company concept employs a once-
through steam generator capable of producing steam having a limited degree
of superheat with an associated heat rate improvement. Otherwise, the
technical approach employed b> all pressurized water reactor manufacturers
is similar with respect to design pressure, use of chemical shim control,
and fuel design.
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Ill-5
The first pressurized water cooled reactor to operate at a sufficiently
high reactor temperature for power generation was the submarine thermal
reactor STR Mark I (now known as SIW), the prototype PWR used in the sub-
marine "Nautilus". The favorable experience with STR led to the full scale
central-station nuclear power project of the Shippingport reactor, which began
operation in 1957. The Shippingport reactor (90 Mwe) pioneered the use of
Zircaloy clad and uranium dioxide (UO2) as a reactor fuel material. These
steps represented two of the more important advances made to date in the
water cooled reactor field.
The Yankee reactor plant, rated at 175 Mwe and located at Rowe,
Massachusetts, represents the first commercial application of a pressurized
water plant in the Dresden I size range. The Yankee plant has operated
successfully for eleven years and as of March 1967 was operating with an
availability factor of approximately 90 percent.
While identical in principle, the current pressurized water reactor
concepts have substantially improved performance characteristics when
compared with the Yankee initial design. These include important advances
in core design performance characteristics, incorporation of chemical shim
and rod cluster control, and optimization of fuel management programs.
The significant design parameters affecting both capital and fuel cycle costs
are similar to those previously listed for boiling water reactor plants, i.e.:
(a) More power delivered per unit weight of loaded uranium,
or greater specific power
(b) More power delivered per unit core volume, or greater power density
(c) Increased exposure of fuel in terms of integrated thermal
energy delivered per unit weight of fuel (MWD/MT)
These improvements in design characteristics result principally from
the employment of chemical shim and rod cluster control, improved fuel
management techniques and heat transfer correlations. A recent PWR
reactor design (1035 Mwe net capacity) has the performance parameters
shown in Table III-2 for comparison purposes.
PWR nuclear power plants each incorporate a closed-cycle pressurized
water nuclear steam supply system and a turbine generator system utilizing
dry and saturated steam. The nuclear steam supply system consists of the
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Ill-6
reactor vessel and its internal components, two or more closed reactor
coolent loops (depending on station size and design) connected in parallel
to the reactor vessel, an electrically heated pressurizer and all necessary
auxiliary systems. The vessel is cylindrical with a hemispherical bottomed
head and a flanged and removable upper head. The reactor core is divided
into three concentric regions. All fuel assemblies are mechanically identical,
but the fuel enrichment for the initial core is different in each region. The
inner region has the lowest enrichment in fissionable uranium-235 and the
outer region has the highest enrichment. This variation results in more uni-
form power distribution throughout the core. A scatter pattern is employed
in the arrangement of the inner region. The fuel rods consist of pellets of
slightly enriched UOg assembled in cold-worked Zircaloy tubes which are
welded closed at the ends.
Each control rod cluster assembly consists of cylindrical absorber rods
fastened together by a spider-type bracket at the top of the cluster. The
control rod clusters provide reactivity control with an adequate safety margin.
The relatively homogeneous distribution of absorber material tends to provide
a more uniform power distribution than the cruciform control used in earlier
PWR's. This system also provides more reactivity control per unit weight
than the cruciform control rods because of the larger surface to volume ratio.
Performance characteristics for future LWR's are projected and
presented in Table III-2. These characteristics are for a future PWR which
was considered to adequately represent LWR status in the analyses to determine
future nuclear energy projections (Section V. B).
2. Advanced Converters
The first attempt to evaluate the future role of advanced converter
reactors in the U.S. civilian power program was initiated at the Oak Ridge
National Laboratory near the end of 1963. Designs generated by sponsors
of various 1000-Mwe concepts were evaluated. These included the spectral
shift control reactor (SSCR), pressurized heavy water-cooled reactors
(HWR-U and HWR-Th), the sodium graphite reactor (SGR), the high temperature
gas-cooled reactor (HTGR), and the molten salt converter reactor (MSCR).
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Ill-7
A pressurized light water reactor (PWR) was also considered for comparative
purposes. The results of these studies indicated that only two of the advanced
converters considered, the HTGR and the MSCR, had the potential of both
lower power costs and better fuel utilization'than the PWR. The remaining
systems were encumbered either with high power costs {SSCR and HWR-Th)
or poor fuel utilization (SGR and HWR-U). On the basis of these early evalu-
ations, it appeared that there was insufficient incentive for continuing the
development of heavy-water-moderated reactors in the U.S. Canadian
studies show that reductions in power costs of HWR's can be obtained by using
organic or boiling-light-water coolants instead of pressurized heavy water.
To verify this, an evaluation of heavy water-mode rated organic-cooled reactors
was undertaken at ORNL in the spring of 1965.
Three concepts were compared: slightly enriched UC-fueled, thorium
oxide-fueled, and thorium metal-fueled systems. The results of the evalu-
ation verified that under the assumed ground rules the HWOCR concepts had
lower power costs and better fuel utilization characteristics than the pressurized
heavy water-cooled reactors and the PWR considered previously. Thus,
by January 1967, three advanced converter concepts--the HTGR, the HWOCR,
and the MSCR--were identified as having better costs and fuel utilization than
light water reactors. The role of these particular advanced converters in the
future U.S. power system still remained unclear, however, because the
evaluations conducted up to that time were limited to a comparis on of converter
reactors with each other, rather than with all reactors that might comprise the
future U.S. power system. The evaluations also did not take into consideration
the effect of the introduction of these converters on the economics of the system
as a whole, the competitiveness of converters with breeders, and the influence
of rising ore prices on the competitive position of converters relative to light
water reactors and breeders.
Based on the degree of continued success of current light water systems,
as well as the availability of low-cost uranium ore and the progress of breeder
reactor development which is being assessed as part of the Commission 's
ongoing civilian power analysis program, it was determined early in 1967
that sufficient data had been developed on heavy water reactor alternatives
to permit readjustment of the program. On this basis, the developmental work
-------
Ill-8
on the heavy water organic-cooled concept has been deferred until the role of
this class of reactors is further clarified. In the meantime, a heavy water reactor
search and development program involving a modest expenditure of U.S. re-
sources is underway. This program, coupled with active international
technical liaison, provides an economical means for the U.S. to maintain
the option to exploit the heavy-water reactors in the future.
HWOCR's are of interest when considering the engineering characteristics
of advanced converters as shown in Table III-3 based on Reference III-3. Five
types of heavy water-moderated organic-cooled reactors (HWOCR) exist due
to fueling type variations:
(a) Slightly enriched uranium carbide
-------
Table III-3 Engineering Characteristics of Advanced Converters
HWOCR
HWBLW
HP3F
EUC
NUC
IvLTK
TUO
TUM
Keactor
Backup
Ref^rence
Fuel type
Uranium
carbide
Uranium
carbide
Uranium tn;tal
(Tn-U)02
Tno riujD-ura-
nium
U02
(Th-'J)02 or
{m-u)c
(Th-U)02 or
(Tn-u)C
?uel configuration
37-pin as-
semblies
19-pm as-
semblies
3 nested cyl-
inders
37-pin
cluster
4 nested cyl-
inders
19-pm as-
semblies
Coated par-
ticles
Coited par-
ticles
Cladding material
SAP
SAP
Ozhennite 0.5
SAP
Ozhennite 0.5
Zircaloy-4
Kone
?ione
Moderator
D20
D20
DjO
D 20
D20
D20
Graphite
Graphite
Coolant
Santov&x 0M plus high
boilers
Boiling
II20
3242
Helium
Helium
¦bjiTLal capacity, Mw
3093
3268
3166
3100
3187
2460
2320
Net electrical capacity, Mw
1057
1118
1070
1076
1048
965
1001
1000
Net station efficiency, %
34.5
34.2
33.8
34,7
32.9
29.8
40.7
43.1
Steam conditions, psig/°F
900/725
900/725
900/685
900/740
600/660
765/514
2400/1000
3500/1050
Coolant temperature, inlet/outlet, °T
590/745
580/750
569/710
605/766
535/665
513/518
757/1449
802/1525
Coolant pressure, psi
285
360
320
280
353
1080
690
695
Coolant flow rate, lb/iir
110 x 106
108 x 106
128 x 106
106 X 106
119 x 106
49.8 X 106
10.3 x 10®
9.3 X 106
Average core power density, kv/liter
16.1
8.1
5.6
12.3
19.6
9.1
7.5
7.0
Average core specific power,
of fuel
24.8
13.6
7.6
26.4
32.2
13.3
60
60
KaxiniaD linear heat rating, kv/ft
23.4
18.4
123
16.5
261
18.7
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Ill-10
(b) The use of zirconium alloys for process tubes for the natural
uranium concepts sintered aluminum product (SAP)
Both UC concepts have SAP-clad fuel pins. However, the uranium metal
fuel element consists of three nested cylinders clad with a zirconium alloy
(Ozhennite 0.5). The reference HWOCR is considered to be fueled with
slightly enriched uranium carbide. However, for the purpose of system
analysis studies, a throwaway fuel cycle core was also proposed for this
concept.
One thorium-fueled HWOCR concept has 37-pin cluster fuel elements
containing ThOg-UOg, and the other has fuel elements consisting of four
nested cylinders of zirconium-alloy-clad thorium-uranium metal. The
conceptual design of a 37-pin cluster ThC-UC-fueled reactor was also pre-
pared by Babcock & Wilcox, but was not included in the ORNL evaluation
because its economic performance was no better than that of the other systems,
and the use of this fuel involved more technical problems. The overall plant
design for the thorium-fueled concepts is basically the same as that for the
Atomic International-Combustion Engineering uranium-fueled reactors,
except that the core design was changed to accommodate the thorium fuels.
The design of the 1000 Mwe HWBLW reactor represents an extra-
polation from the Canadian CANDU-BLW-250 Mwe reactor, which is smaller
in overall size but in many other important respects is essentially the same as
the 1000 Mwe reactor. The BLW-250 reactor is being designed and built by
Atomic Energy of Canada, Ltd.
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Ill-11
The HTGR backup design is essentially a scaleup of the 330-MWe
Fort St. Vrain plant. This concept consists of a graphite-moderated helium-
cooled reactor operating on the thorium-U-233 fuel cycle. The graphite serves
as the moderator, reflector, and fuel-bearing structure, and the fuel is in
the form of carbon-coated particles of UC^ + ThC^ or UC2 + TK^. The core
and complete helium circulating system are housed in a prestressed-concrete
reactor vessel.
The HTGR reference design includes the following improvements over
the backup design-
fa) On-line refueling
(b) A prestressed concrete pressure vessel
(c) A radial-flow steam generator, and
(d) Supercritical steam conditions
In both the backup and reference HTGR's, the initial operation of the reactor
is normally with uranium highly enriched in U-235 as fissile material. Bred
uranium is recovered from the spent fuel and is recycled along with enough
fully enriched uranium to maintain criticality. In the typical equilibrium cycle,
the fuel elements have a mixture of two types of coated particles, makeup
particles containing only highly enriched U-2 35 and fertile particles containing
the thorium and all the recycled uranium. This permits separation of the
bred material from makeup material.
3. Thorium Systems
There is an overlap between the advanced converter and thorium systems
classifications. The HWOCR and HTGR advanced converters that utilize
thorium fuel types can also be classed as thorium systems. However.this section
will consider only the remaining types, molten salt reactors.inthe following dis-
cussion. Power reactor types employing thorium are discussed to considerable
detail in Reference III-4.
Molten salt technology has been studied extensively at ORNL since 1950.
There have been two molten salt reactors--the Aircraft Reactor Experiment
in 1954 and the Molten Salt Reactor Experiment (MSRE)--as well as a broad
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Ill-12
base of related applied research in this concept and other fluid-fuel
reactors. These experimental reactors provided a varied background of
experience in complete circuits of circulating fuel, including reactor kinetics
response, pumping of fluid fuels, heat removal, and remote maintenance.
Since it achieved critically in June 1965, the MS RE operated successfully
until placed on standby m December 1969, mostly at a power level of 8. 0 Mwt.
This operation has served to demonstrate the following important design
features of the experiment-sized single-region molten salt concept:
(a) The practicality of high temperature (1200 F) operation
of a molten salt fuel
(b) The sustained performance of basic system components,
such as pumps, heat exchangers, and instrumentation,
with molten salt fuel
(c) Satisfactory performance of remote maintenance
(d) Removal of xenon and other volatile fission products from
the molten salt
(e) On-line refueling and fuel adjustment, and
(f) Self-regulation and good response to changes in power
demand
Preliminary reactor designs, including the 1000 Mwe MSBR as well
as an advanced converter, are currently under investigation. Program plans
include:
(a) Demonstration of dimensional and structural stability of graphite
during long exposure to fast-neutrons
(b) Establishment of long term compatibility of Hastelloy N in
the molten salt and neutron environment
(c) Development of remote maintenance equipment
(d) Removal of fission products and Pa-233 from molten salts
during reactor operation
(e) Scale-up of system components, especially the pumps and heat
exchangers
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Ill-13
As in all reactor development programs, there is a difficult transition
from an experimental facility such as the MS RE to a large scale commercial
plant such as the MSBR. This concept has not yet received significant
industrial or utility support, and major R&D efforts will be required to
develop the concept commercially.
Previously, the reference design for the development of the MSBR has
been the ORNL two-region, two-fluid system with fuel salt separated from
the blanket salt by graphite tubes. The fluids consisted of lithium and
beryllium fluorides containing UF^ and ThF4 for the fuel and blankets materials,
respectively. The on-site fuel reprocessing employs fluorides-volatility and
vacuum distillation operations for the fuel steam and direct protactinium re-
moval for the blanket steam. This reference design was assessed by the
Thorium Task Force and was the basis for the Systems Analyses Task Force
overall assessment effort.
Graphite irradiation experience has shown that dimensional changes
can occur which result in an initial volumetric contraction followed by expan-
sion. The rate of expansion, after the initial contracted volume is attained, in-
creases with increasing exposure so that eventually the expansion limits the useful
life of the graphite. In addition, the factors which control the lifetime dosage
are graphite strength and changes in pore structure under irradiation.
A consequence of the irradiation experience was the further reassess-
ment of the MSBR development effort due to the considerable uncertainty as
to the practicality of using graphite as a structural material to separate
fluids in the reference two-fluid MSBR concept. Simultaneously chemical
research results indicated that molten salt reactors potentially could be
operated economically as single fluid systems. These developments were
associated primarily with the evidence that protactinium as well as rare earth
fission products could be separated from single fluid salts. Thus in mid 1968,
a single fluid, two-region MSBR concept was proposed, and a preliminary
conceptual design prepared in which the graphite no longer serves as a
structural material to separate two distinct fluids, but primarily serves as a
moderator and a separation medium for two fuel regions of a single fluid.
An important consideration in the new design was theoretical and preliminary
experimental evidence that the U-233, and possibly rare earth fission products,
-------
Ill-14
could be separated from a mixed thorium uranium fuel salt by reduction
extraction employing liquid bismuth. This, combined with nuclear con-
sideration of the single fluid design, indicated that fuel breeding gains and
economics comparable to the reference two fluid system could be achieved
by the proposed single fluid concept.
An alternative for the MSBR is provided by the Molten Salt Converter
Reactor (MSCR), a single-region, single-fluid reactor moderated by graphite,
which is essentially the same as the single-fluid MSBR except that the fuel
is processed on a much longeri processing cycle. Thus, an MSCR can be
converted to an MSBR by appropriate installation of processing equipment.
The MSCR is a reactor which utilizes fluoride volatility and vacuum distillation
processing.
Performance characteristics for the MSCR and the MSBR are summarized
in Table III - 4. Uranium-thorium and plutonium-thorium fuel cycles can be
utilized in MSR's.
4. Breeder Reactors
Nuclear power reactors of the breeder type (Ref. Ill -5) produce more
nuclear fuel than they consume. Thus they would make it feasible to utilize
enormous quantities of low-grade uranium and thorium ores dispersed in
the rocks of the earth as a source of low-cost energy for thousands of years.
In addition, these reactors would operate without adding noxious combustion
products to the air. It is in the light of these considerations that the U.S.
Atomic Energy Commission, the nuclear industry, and the electric utilities
have mounted a large-scale effort to develop the technology whereby it will
be possible to have a breeder reactor generating electric power on a com-
mercial scale by 1984.
Nuclear breeding is achieved with the neutrons released ty nuclear
fission. The fissioning of each atom of a nuclear fuel, such as uranium 235,
liberates an average of more than two fast (high-energy) neutrons. One of
the neutrons must trigger another fission to maintain the nuclear chain
reaction; some neutrons are nonproductively lost, and the remainder are
available to breed new fissionable atoms, that is, to transform "fertile"
isotopes of the heavy elements into fissionable isotopes. The fertile raw
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Ill-15
TABLE lU-4. SUMMARY OF DESIGN AND PERFORMANCE
PARAMETERS FOR MOLTEN SALT REACTORS
Molten Salt Molten Salt
Converter Breeder
Core height, feet 20.8 13.7
Core diameter, feet 16.6 9.7
Blanket thickness, feet 2.0
Core power, Mwt 2222 1812
Blanket power, Mwt 410
Average core power density, Kw/1 17 64
Graphite replacement life, years 2. 1
Specific fuel inventory, Kg/Mwe 1.63 1.0b
Breeder ratio 0.96 1.07
Annual fuel yield, %/year 4.8
Fuel doubling time, years 14. 4
Thermal to electrical conversion
efficiency, % 45,0 45.0
-------
111-16
materials for breeder reactions are thorium-232, which is transmuted into
uramum-233, and uranium-238, which is transmuted into plutonium-239.
Breeding occurs when more fissionable material is produced than is
consumed. A quantitative measure of this condition is the doubling time: the
time required to produce as much net additional fissionable material as was
originally present in the reactor. At the end of the doubling time the reactor
has produced enough fissionable material to refuel itself and to fuel another
identical reactor. An efficient breeder reactor will have a doubling time in
the range of from seven to 10 years.
Two different breeder systems are involved, depending on which raw
material is being transmuted. The thermal breeder, employing slow neutrons,
operates best on the thorium-232-uranim-233 cycle (usually called the thorium
cycle). The fast breeder, employing more energetic neutrons, operates best
on the uranium-238-plutonium-239 cycle (the uranium cycle). Nonproductive
absorption of neutrons is less in fast reactors than it is in thermal reactors,
resulting in a decrease in the doubling time. For this reason, only fast breeder
reactors are discussed below.
The breeder reactor types are defined based on the type of coolant
employed to carry off the heat of fission and deliver it to a power generating
system. The coolants proposed were water and molten salts for thermal
breeding and inert gas, liquid metal, and steam for fast breeding. The molten
salt breeder reactor (MSBR) was discussed in the previous section. As shown
in Table III-1, the remaining breeders (LMFBR, SCFR, and GCFR) make up
the breeder reactor class. The liquid-metal-cooled fast breeder reactor is
discussed further in the following section. The SCFBR and the GCFR are
discussed under the heading "Alternate Coolant Fast Breeder Reactors. "
a. Liquid-Metal-Cooled Fast Breeder Reactor (LMFBR). In the U.S.
and several other countries decisions were made that a fast breeder reactor
cooled with liquid metal was an attractive concept to develop. Several design
features of the LMFBR are of interest. The core of a fast reactor can be quite
small. For economic reasons the reactor must be operated at a much higher
power density than ordinary fission reactors. The active core volume is
therefore only a few cubic meters and is roughly proportional to the power
output. The power density is approximately 400 Kw/liter.
-------
Ill-17
In order to carry off the heat while maintaining the fuel at a
reasonable temperature, sodium must flow through the core at a rate of tens
of thousands of cubic meters per hour. To provide channels for the flow of
sodium, the fuel is divided into thousands of slender vertical rods.
The fuel is preferably in a ceramic form such as oxide or carbide.
These ceramics are stable during long exposures to heat and radiation, have
very high melting points, and are relatively inert in liquid metal. The
fissionable component of the fuel can be enriched uranium-235, plutonium-239,
or a mixture of the two. Performance characteristics of LMFBR's utilizing
various fuel forms are shown in Table in-5.
Much of the breeding takes place in the blanket that surrounds the
fast reactor core, The blanket consists of uranium-238 in stainless steel
tubes. Since there is a certain amount of fission in the blanket, it too must
be cooled by the sodium.
Interspersed through the core region are numerous rods with
safety and control functions. They maintain the power output at the desired
level and provide the means for starting and stopping the reactor. The rods
are filled with neutron absorbing material such as boron carbide or tantalum
metal. Fast breeder reactors require fewer control rods than thermal reactors.
Three major reactors will carry the burden of the A. E.C. 's
program to develop an LMFBR. Two of them are already in operation:
(1) The Experimental Breeder Reactor II (EBR-II),
since 1964
(2) The Zero Power Plutonium Reactor (ZPPR),
since 1969
The third reactor, Fast Flux Test Facility (FFTF), is being designed on
the basis of data obtained from EBR-II, ZPPR, and smaller facilities.
EBR-II is a fast-neutron test reactor operated by the Argonne
National Laboratory at the A. E. C. 's National Reactor Testing Station m
Idaho. This reactor is the focal point of the program for testing fuels and
irradiating materials for the LMFBR. At mid-1971, EBR-II has produced
more than 1, 000, 000 Mw-hr of power (thermal) and over 250 x 10 Kw-hr
electricity. It has achieved its design power of 62. 5 MWt.
-------
TABLE HI-5. PERFORMANCE CHARACTERISTICS OF
LIQUID-METAL-COOLED REACTORS
Reference
Oxide
Advanced
Oxide -
Negative
Sodium
Advanced
Oxide-
PosiUve
Sodium
Oxide
Converter
Reference
Carbide
Advanced
Carbide
Ca rbide
Converter
Thermal power, Mw 2197
Net electrical capacity, Mw 880
Average core power density, aKw/liter
Average core specific power, aMwt/MT 175
Core fissile plutonium inventory. Kg
Total fissile plutonium inventory, 3430
Net gain of fissile plutonium, Kg/yr
Core average burnup, Mwd/MT 80, 000
Breeding ratio'' 1.27
Exponential doubling time, cyeara 15.00
1975
860
230
2410
101, 000
1. 34
8.00
2103
915
230
2100
97, 000
1. 31
7.00
2162
910
106. 4
3800
80, 000
852
2300
850
113
2410
79, 360
1.45
8. 6
2295
900
167. 9
1300
110,300
1. 50
4. 8
20 i2
860
126.7
3250
80, 000
. 89
aCalculated as linear averages over equilibrium cycle.
^Defined as fertile captures divided by absorptions in fissile plutonium
cBased on one-year out-of-core holdup of fissile plutonium.
Reference: WASH 1098, Tables E-5 and E-6.
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Ill-19
The Zero Power Plutonium Reactor is also operated by Argonne
National Laboratory. It has the size and a large enough inventory of plutonium
(s 3000Kg) to allow full-scale mockups of the plutonium fuel arrangements
that will be used in the large commercial breeders envisioned for the 1980's
and beyond.
The Fast Flux Test Facility will operate at a very high neutron
flux to produce the radiation effects on fuel and structural materials that
will take place in a commercial breeder reactor. It will operate at 400 Mwe
power and will be built at the A.E.C. 's site in Richland, Washington. Initial
operation is anticipated in 1974.
b. Alternate Coolant Fast Breeder Reactors. Most of the development
work of fast breeder reactors has been based on the use of liquid metal cooling,
but in the early 1960's some interest was expressed in the use of other coolants
as a means of alleviating some of the problems associated with liquid metals.
Numerous coolants were considered. However, most were eliminated either
by chemical or metallurgical evaluations (air, hydrogen, carbon monoxide) or
heat transfer considerations (neon, argon). In addition to relatively poor heat
transfer, argon also has problems of neutron activation, and neon is very
expensive. Nitrogen has a high neutron-absorption cross section and might
cause nitriding at high temperatures. Supercritical SOg cooling has been
studied at the Oak Ridge National Laboratory and appears to be feasible.
However, experience with this system at the required operating conditions
is very limited. Thus, the list of alternate coolants of interest for fast breeder
reactors was reduced to carbon dioxide, helium, and steam. Although from
purely thermal and cost standpoints there appears to be little choice between
COg and helium, CC^ cooling leads to higher core pressure drops, higher
fuel ratings, and greater potential for flow-induced vibrations. Thus, present
designs of alternate-coolant fast breeder reactors are based on the use of
helium or steam cooling. These reactor types are discussed in depth in
Reference III-6.
Steam-cooled fast breeder reactors have been studied in varying
degrees of depth since 1961. At least four groups in the United States have
studied the concept, and studies have also been made in Germany by the
-------
Ill-20
Karlsruhe Nuclear Research Center, Institute of Reactor Development; in
Sweden by A.B. Atomenergi; and by the Euratom-Belgium Fast Reactor
Association.
The study reported in 1965 by the Babcock & Wilcox Company
jointly with American Electric Power Service Corporation was based on use
of supercritical steam as a coolant and, although the concept appeared to be
economically competitive, the breeding ratio was only about 1. 03. It was
concluded that the concept was not consistent with the objectives of the AEC
breeder reactor program but that it did fulfill the requirements for an ad-
vanced converter. The low breeding ratios found in this study, and the
conviction that the core could be cooled with low-pressure steam, led the
Babcock & Wilcox Company to propose at the Argonne National Laboratory
in October 1965 that a lower steam pressure would result in a higher breeding
ratio. Their paper indicated that, with a steam pressure of 1200 psi, it would
be possible to attain an overall breeding ratio of about 1. 40. They had not
developed a complete design based on the parameters of the lower pressure
steam-cooled reactor. Thus in 1966 and 1967, under contract to the Oak
Ridge National Laboratory, Babcock & Wilcox Company and Sargent & Lundy
Engineers developed a more complete design study of the low pressure con-
cept for use by the Alternate Coolant Task Force in the evaluation of steam -
cooled breeder reactors. A rather complete study by the Karlsruhe group
was published in August 1966 based on the use of steam at a pressure of
approximately 2650 psia. The German study showed a breeding ratio of about
1.15 and a doubling time of about 37 years. Thus, considerations of the range
of steam pressures from 1250 to 3700 psia was available to the task force
in three design studies. The Oak Ridge National Laboratory attempted to
normalize design parameters of these three studies sufficiently to evaluate
the potential of steam cooling for fast breeder reactors, and to obtain some
understanding of the economics, physics, safety, unsolved problems, and the
effect of steam pressure on the concept. As a supplement to this evaluation,
the Pacific Northwest Laboratory was asked to perform a parametric study of
reactor design over the range of steam conditions of current interest.
Three steam-cooled fast breeder reactors (SCBR) have been con-
sidered in the evaluation (Ref. III-6). Some of their performance characteristics
-------
Ill-21
are shown in Table III -6. The first is a high pressure concept cooled by
steam which enters the reactor at 3700 psia and 750° F. Steam enters the
turbine at 330 psia and 994° F, with reheat to 950° F. This high pressure
SCBR concept was developed jointly by the Babcock & Wilcox Company and
American Electric Power Service Corporation.
An intermediate pressure SCBR concept developed by the Karlsruhe
group is the second concept considered. In this system, the steam enters the
reactor vessel at 2680 psia and 710° F and enters the turbine at 2350 psia and
a temperature of 1000° F. The steam is reheated in a surface-type heat
exchanger to 950° F.
The third SCBR considered was developed by Babcock & Wilcox
Company and Sargent & Lundy Engineers and is referred to as the low-
pressure SCBR design. In this concept steam enters the core at 1250 psia and
575° F, and the turbine throttle conditions are 1050 psia and 925° F. There is
no reheat of the steam in this power cycle.
In all three SCBR concepts, the steam used for cooling the reactor
core is generated in a Loeffler-type cycle that uses steam from the reactor
exit as a heat source. Steam generated by mixing the high-temperature exit
steam with feedwater is circulated through the reactor by a steam circulator.
In the low pressure design, the boiler and circulator are integral with the
reactor vessel so that piping between the Loeffler boiler, the steam circulator,
and the reactor is eliminated. In the other two concepts, the steam generators
and circulators are located separately from the reactor vessel and are connected
by piping.
Conceptual design studies of the GCFR were started in 1961 by the
General Atomic Division of General Dynamics Corporation. These studies
were continued and expanded in additional work by Gulf General Atomics under
an AEC contract starting in 1963. A conceptual design study to establish the
practicality of a 1000 Mwe GCFR power plant was undertaken in August 1965
as a joint effort by GGA and the East Central Nuclear Group (ECNG). The
results to date of the ECNG study were made available to the AEC and the
Alternate Coolant Task Force. This study was updated and reoptimized for
the current assessment of civilian power reactors under contract to ORNL,
and the results were published in a report by GGA.
-------
Table III-6 Performance Characteristics of Alternate Coolant Past Breeder Reactors
Steam-Cooled .Reactora Gas-Cooled Reactors
1250-psi 2680-psi 3700-psi Derated Reference Carbide-
Plant Plant Plant Plant Plant Fueled Plant
Thermal power, Mw
2,900
2,519
2,326
2,681
2,530
2, 67S
Net electrical capacity, Mw
1,012
1,000
970
1,000
1,000
1,000
Average core power density,a kw/liter
355 *
285
445
245
265
545
Average core specific power,a
Mv/kg fissile
0.74
0.66
0.94
0.80
0.82
1.30
Core fissile plutonium inventory,a kg
3,555
3,530
1,980
2,987
2,727
1,688
Total fissile plutonium inventory,a kg
A, 085
3,835
2,325
3,797
3,523
2,520
Net gain of fissile plutonium, kg/yr
318
95
65
349
356
450
Core average burnup, Mwd/MT
70,000
57,500
58,000
74,800
75,000
112,200
Breeding ratio*1
1.38
1.14
1.11
1.48
1.48
1.60
Exponential doubling time,c years
12
38
41
11
10
5.5
Fuel life, years at a power factor of 0.8
3.0
2.65
1.6
2.5
2.5
2.5
Calculated as linear averages over equilibrium cycle.
^Defined as fertile captures divided by absorptions in fissile plutonium.
Based on one-year out-of-core holdup of fissile plutoniua.
H
n
n
I
ra
rv>
-------
111-23
Incentives for developing the helium-cooled fast reactor (GCFR)
include the attainment of a high breeding ratio in a system utilizing gas
cooling technology. The breeding ratios obtained in the GCFR can be equal to
or better than those of the LMFBR. Principal components, including the
helium circulators, steam generators, prestressed concrete pressure vessel,
and helium piping, would draw heavily on technology for existing thermal
gas-cooled reactors and on current HTGR development programs. Fuel
element design would utilize the current and planned AEC development programs
for sodium-cooled fast reactors. The high internal breeding ratio of the GCFR,
along with the provision for on-load variation of flow orificing, would permit a
fuel cycle in which the entire core was reloaded at one time. This would mini-
mize the frequency of outages for refueling.
Three helium-cooled designs were also considered:
(1) A derated oxide fueled design (GCFR-4D) with a
limit on the cladding temperature set 50° C below
that specified in the reference design
(2) A reference oxide fueled design (GCFR-4) with
relatively high values for maximum pin heat
rating and cladding temperature
(3) A carbide fueled design operating at higher pin rating,
gas pressure, power density, and specific power
Some performance characteristics for these breeder reactors are shown in
Table ffl-6.
In a typical GCFR, the reactor and steam generators are contained
in the prestressed concrete pressure vessel which also serves as the biological
shielding. Electricity is generated from a high pressure steam cycle with a
net thermodynamic efficiency of almost 40 percent. The steam generators
are supplied with 1250- to 1750-psi high-temperature helium. The coolant flow
is downward through the core, with the fuel elements cantilevered from a deep
top-mounted grid plate. Each fuel element is comprised of metal clad mixed
uranium-plutonium oxide or carbide fuel rods. Overall conversion ratios of
1.45 to 1. 60 are obtained, and the internal conversion ratios are about 1. 0.
-------
Ill-24
B. Fusion Reactor Power
The generation of nuclear power by controlled nuclear fusion can
probably be achieved. An optimistic projection is that fusion power will be
available in appreciable quantity by the year 2000. Controlled fusion research
has passed through several epochs, the first of which was initiated by four
items. First came measurements of reaction energies and rates between
hydrogen isotopes and other light elements, which showed that under proper
conditions large energy releases would be possible. Second, the well-known
laws of single particle physics seemed to show how an assembly of high energy
ions and electrons could be confined in magnetic fields long enough to establish
the proper conditions. Third, the radioactive ingredients and by-products of
fusion appear to be much less hazardous than those associated with nuclear
fission; therefore, fusion reactors would be simpler and safer than fission
reactors. Fourth, deuterium is a fusion fuel in plentiful supply--one part in
7000 of ordinary hydrogen, and extraction from ordinary water is not difficult.
So matters stood in the early days, say up to 1955.
Under the present scientific development program various schemes
are being studied and postulated to confine the fusion plasma. The main
schemes being developed so far involve use of large volumes of high magnetic
fields. Plasma ions and electrons are hindered by magnetic forces from moving
across the direction of magnetic fields, but can spiral along the field lines.
Thus (naively), confinement in the two directions perpendicular to the field
direction is achieved, and one might have to worry only about confinement
along the field direction. The various devices which may lead to a feasible
fusion power reactor are:
(1)
Stellarator
(2)
Tokamak
(3)
Internal conductor devices
(4)
Fast pulsed (0 - pinch)
(5)
Hot electron mirror
(6)
Ion injection mirrors
-------
Ill-2 5
(7) Astron {a mirror device)
(8) Continuous flow pinch, and
(9) Laser ignition
The description and status of these various devices are presented by Rose in
Reference III-7. At this time hopes are high that the Tokamak device will
establish the feasibility of a fusion reactor. It utilizes a toroidal-shaped
plasma which is also the secondary loop of a transformer.
Problems must be solved before a controlled fusion reactor can be
attained. There are some problems that are substantially independent of the
particular geometric model or device:
(1) Plasma conditions in imagined practical devices, such as ion
and electron temperatures, the fraction of fuel burned up per
pass through the reactor, and radiation from the plasma surface
must be calculated. However, a major difficulty with all calcu-
lation of electron and ion temperatures with increasing fractional
fuel burnup is that the rather restrictive assumptions used, i.e.,
no turbulence or space charge effects in the plasma, render the
calculations nebulous. A subfield of fusion plasma engineering
needs developing before a fusion reactor can be sensibly designed.
(2) Regenerating tritium (for a Deuterium-Tritium reactor) in a
surrounding moderator-blanket by means of the 14.1-Mev
neutrons in a Li reaction seems adequately assured by using a
liquid Li or Li salt coolant with a neutron moderator between
the blanket and vacuum walls of the reactor.
(3) Heat deposition, temperature of the moderator and vacuum
wall, and heat removal must be determined since heat deposition
(and removal) per unit volume on the vacuum wall facing the
plasma determines the power capability of the whole system.
(4) Providing large quantities of high magnetic field and structure
to withstand high stress is a problem. For example, a simple
solenoid generating 150, 000 gauss has a magnetic bursting force
of 900 atm. on its windings.
-------
Ill—2 6
(5) The effects of material radiation damage to the vacuum wall
by the 14* 1-Mev neutrons must be minimized since the conse-
quences of such damage may be frequent and expensive
replacement of much of the structure.
(6) Size and cost, which are implicit in many of the above are, of
course, the major factors in determining the feasibility of a
fusion reactor. Size is large for lowest power cost and will
increase in the future so that 10, 000 Mwt is liable to be quite
acceptable in 2000. Cost estimates indicate thatcostmight be
$6-20 per Kwt, an attractive cost range. Other problems
are model-dependent; some device concepts seem to require
additional developments.
These problems can be described and a physical feel can be obtained
with the aid of Figure III-1, a fusion reactor schematic consisting of a series
of concentric cylinders.
Figure III -1. Schematic of Controlled Fusion Reactor
-------
Ill-27
The main confining magnetic field is into (or out of) the paper; whether the
cylinder is the center section of a stabilized mirror device or is wrapped
into a torus need not concern us here. The fusion plasma occupies the
evacuated center, is surrounded by a neutron-moderating blanket and, at
large radius, by a set of magnetic field coils.
Since its containment is imperfect, some of the plasma fuel is continually
being lost from the ends or sides. Thus, it is being replaced by some injection
process into the center. A certain throughput of plasma is needed to keep up
its density. The plasma is at least partly heated by its own reaction.
For a deuterium-tritium reactor, tritium must be regenerated. The
general approach, then, in Figure III-1 is to make the vacuum wall and blanket
supporting structure of thin section refractory metal. Within it, there would
be liquid lithium or a lithium salt coolant, plus an artfully disposed neutron
moderator (probably partly of graphite). Leading choice for metals is
niobium in that it can be formed and welded, retains its strength at 1000°C,
and is transparent to tritium. This transparency helps in two ways:
tritium generated in the lithium-bearing coolant is not trapped in the metal;
and tritium can be recovered by diffusion through thin section walls into
evacuated recovery regions. Some additional neutrons also come from the
niobium via (n, 2n) reactions, but in this particular respect molybdenum
would be a better material. The tritium fuel doubling time in a fusion reactor
might be less than one year. This short doubling time for fusion is contrasted
to the longer one (a* 20 years in some designs) for fission breeder reactors.
The approximate size of the fusion reactor can be estimated. Fairly
simple nuclear calculations establish that the blanket plus a radiation shield
(not shown) to protect the outer windings must be 1.2 to 2.0 m thick. This
substantial thickness implies not only substantial blanket cost, but also very
high magnetic field cost, to energize such a large volume. The only way to
make the system pay is to have it generate a great deal of power, but nearly
all this power must pass from the plasma into or through the vacuum wall.
Engineering limits of power density and heat transfer then dictate large
plasma and vacuum wall radii as well--between 1 and 4 m, say. Then overall
size will be large, and total power will be high—almost certainly more1 than
1000 megawatts (electric) and perhaps 5000 megawatts.
-------
111-28
Energy is deposited in the vacuum wall facing the plasma, mainly from
three sources:
(1) Some of the fusion neutrons suffer inelastic collisions as
they pass through
(2) Gamma rays from deeper inside the blanket shine onto the
back side
{3} ALL electromagnetic radiation from the plasma is absorbed
there
The plasma itself makes no additional load, being imagined to be pumped out
elsewhere. The three sources may constitute 10 to 20 percent of the total
reactor power. This is a modest fraction; but the vacuum wall region is thin,
and heat depostion (and removal) per unit volume determines the power
capability of the whole system.
The size of the magnetic field windings generating 150, 000 gauss causes
a problem with the stresses that arise resulting from the magnetic bursting
force. Almost all conceptions involve superconducting coils at 4°K, or at
least cryogenically cooled ones at 10° to 20°K. This is the reason for placing
them outside the blanket, outside a radiation shield; otherwise the refrig-
eration problem would be intolerable. To make a reinforcing structure for
operation at such a temperature, with the size and stress loads described, is
a task yet to be fully contemplated. Titanium is very strong at such low
temperatures; but it is also very brittle--as are most other materials under
those conditions.
Neutron damage is a very serious problem, for either a fission or
fusion reactor. In one way, fusion appears at a substantial disadvantage,
as follows. One fission reaction produces 200 Mev and about 2.5 neutrons,
each with no more than about 2 Mev. One fusion reaction produces 17. 6 Mev,
of which 14.1 Mev appears in one high energy neutron. Thus, the "energetic
neutrons/watt" is an order of magnitude higher in fusion than in fission, and
the structural damage caused by these neutrons is correspondingly high.
-------
m-29
REFERENCES
III-1. "Potential Nuclear Power Growth Patterns, " WASH 1098,
U. S.A.E.C. Division of Reactor Development and Technology,
December 1970.
IU-2. "Current Status and Future Technical and Economic Potential
of Light Water Reactors," WASH 1082, U.S.A.E.C. Division
of Reactor Development and Technology, March 1968.
HI 3. "An Evaluation of Advanced Converter Reactors, " WASH 1087,
U.S. A.E.C. Division of Reactor Development and Technology,
April 1969.
Ill-4. "The Use of Thorium in Nuclear Powe r Reactors, " WASH 1097,
U.S.A.E.C. Division of Reactor Development and Technology,
June 1969.
III-5. Seaborg, G. T. and J. L. Bloom, "Fast Breeder Reactors. "
Scientific American, Vol. 223, No. 5, November 1970,
pp. 13-21.
ni-6. "An Evaluation of Alternate Coolant Fast Breeder Reactors, "
WASH 1090, U.S. A.E.C. Division of Reactor Development and
Technology, April 1969.
Ill-7. Rose, D. J., "Controlled Nuclear Fusion: Status and Outlook, "
Science, Vol. 172, No. 3985, May 21, 1971, pp. 797-808.
-------
IV-1
IV. EXISTING AND PLANNED GENERATING STATUS
BY TYPE AND REGION
The existing and near-term planning status of central station nuclear
plants is presented for each census region in Tables IV-1 through IV-9.
This status is defined as encompassing those units operable, under con-
struction, or planned in the near future. For each state, the nuclear plant
name is given. The net plant power capacity in Mwe and type for each unit
are listed in the year projected as being the startup time. In the latter
years, such as those spanning the period from 1976 through 1980, additional
new plant purchases can be expected. The total electrical capacities for a
given year are shown after summing the individual capacities over the
particular census region. The data in these tables were compiled from
Refs. IV-1 through IV-3. More recent information for update purposes
was obtained from U. S.A.E.C. immediate releases and recent issues of
Nuclear News and Nuclear Safety. Therefore, the tables are considered to
be current through September 1971.
The geographical distribution of nuclear power plants in the United
States is presented in Figure IV-1 (Ref. IV-2). This figure includes one
operable and two planned dual purpose plants which generate electrical
power.
-------
TABLE IV-1. CENTRAL STATION NUCLEAR PLANTS
CENSUS REGION- NEW ENGLAND
(Maine, New Hampshire, Vermont, Massachusetts, Rhode Island, Connecticut)
State
Me.
N.H.
Vt.
Mass.
R.I.
Conn.
Nuclear
Plant Name
Me. Yankee
Seabrook
Vt. Yankee
Pilgrim
Yankee -Rowe
(None)
Conn. Yankee
Millstone 1
Millstone 2
Region Total
Pre-
1970
1970
<1961) 175
(PWR)
(1967)575
(PWR)
1971
514( PW R)
1972 1973
790(PWR)
655(BWR)
6S2(BWR)
1974
1975
1976
1977
1978
885
197 9
1080
750 652
514
1445
828(PWR)
828 0*
0*
0*
885*
•New purchasee may be expected.
<
i
to
-------
TABLE IV-2
CENSUS REGION: M1UDLK ATLANTIC
(Naw York, Ne«rjeidcy, Pennsylvania)
Pre-
1970
(1963)265
(BWR)
Nuclear
State Plant Name
N.Y. Bell (Inactive)
F itzpatnck
Indian Point 1
Indian Point 2
Indian Point 3
Indian Point 4
Indian Point S
Shoieham
Girina
Nine Mile Pt.l
N ine Mile Pt. 2
N.J. Forked River
Newbold Island 1
Newbold Island 3
Salem 1
Salem 2
Oyster Creek (1989)560
(BWR)
Pa. Limerick 1
Limerick 2
Peach Bottom 1
Peach Bottom 2
Peach Bottom 3
Three Mile
Island 1
Three Mile
Island 2
Pa.Pwr. iLt. 1
Pa. Pwr. lcLt.2
Beaver Valley X
Beaver Valley 2
Shlppingport
1670
1971
1972
873(PWKJ
1975 1976
1977
1978
838(BWR)
1979
19B0
96S(PWR)
420(PWR)
SOO(BWR)
ioso(pwe)
lllS(PWB)
UIS(BWR)
819(BWR)
lOSB(BWR)
10S5(BWR)
(1967) 40
(GCGM)
1115(BWR)
UOO(BWR)
mo(pwa)
1088(BWR)
106fi(BWB)
108MBWR)
83KPWR)
lOSS(BWR)
90?(PWR)
847
-------
TABl.E IV-3
CENSUS REGION- SOUTH ATLANTIC
(Delaware, D.C., Florida, Georgia, Maryland, N. Carolina, S. Carolina, Virginia, W. Virginia)
;>«uclear
State
Plant Name
Del.
(None)
D.C
(None)
r la.
C i y stal River 3
Cr\ stal River 4
Hutchinson Isi. 1
Hutchinson Isl. 2
Turkey Point 3
Turkey Point 4
Ga.
Hatch 1
Match 2
Hancock 1
Hancock 2
Aid.
Calvert Cliffs 1
Calvert Cliffs 2
N.C.
Brunswick. I
Brunswick 2
TUcGiure 1
McGuire 2
Uske County 1
Wake County 2
Wake County 3
Wake County 4
Wilmington X
s.c.
Robinson 2
Oconee 1
Oconee 2
Oconee 3
Parr 1
Va.
Noi th Anna 1
Not th Anna 2
Surry 1
Surry 2
North Anna 3
North Anna 4
W. Va.
{None)
Region Total
Pre-
1870
1970
1971
693
800(PWR)
786IBWR)
1978
897
UOO(BWR)
1979
1980
1100
7B0(PWR)
780(PWR)
3997
BB6(PWR)
845(PWR>
845(PWR>
900
940
940
2517
3311
1933
2407* 3843* 3852* 2015*
915*
•New purchases to be expected.
HH
<
I
-------
TABLE IV-4
CENSUS REGION- EAST NORTH CENTRAL
(IlUnola, Indiana, Michigan, Ohio, Wisconsin)
State
Nuclear
Plant Name
1970
1971
1972
1973
1974
1975
1976
1977
197B
1979
1980
111.
Dresden 1
Dresden 2
Dresden 3
LaSalle 1
La Salle 2
Quad Cities 1
Quad Cities 2
Zion 1
Zion 2
(1960)200
(BWH)
BOtHBWR)
809IBWR)
809(BWR)
809(BWR)
10S0(PWR)
107MBWR)
1078(BWR]
10S0(PWR)
Ind.
Mich.
Ohio
Wise.
Bailly 1
Cook 1
Cook 2
Fermi 2
Midland 1
Midland 2
Palisades
Big Rock PI.
Fermi
Davis-Besse
Zimnier 1
Zimmer 2
Kewaunee
Pi. Beach 1
Pt. Beach 2
LaCrosse
(1962)70
(BWRJ
<1966)61
(SC. F)
10M(PWR)
loeo(pwR)
11>3(BWR)
S60(BWR)
492
700(PWR)
81S(PWR)
872
-------
TABLE IV-5
CENSUS REGION EAST SOUTH CENTRAL
(Alabama, Kentucky. Mississippi, Tennessee)
Nuclear
State
Plant Name
Ala.
Browns Ferry 1
Browns Ferry 2
Browns Ferry 3
Farley 1
Farley 2
Ky.
<.None)
Miss.
(None)
Tenn.
Sequoyah 1
Sequoyah 2
Umt 6
Unit 7
Unit B
Unit 9
Watts Bar 1
Watts Bar 2
1970
19T1
1972
1065(BWR)
1065(BWRJ
1973
1065CBWR)
1974
1975
828(PWR)
1976
1977
829{PWR)
1978
1979
1980
1 124(PWR)
1124(PWR)
1150(PWR)
1170(PWH)
1150(PWR)
1230(PWR)
1170( PWRJ
1330(PWH)
Region Total
2130
1065
1246
829
2320*
4379*
1230*
0*
0*
~ New purchases may be expected.
<
I
cr>
-------
TABLE IV-6
CENSUS REGION: WEST NORTH CENTRAL
(Iowa, Kansas, Minnesota, Missouri. Nebraska, N. Dakota, S. Dakota)
State
Nuclear
Plant Name
Iowa
Arnold
Kan.
(None)
Minn.
Prairie Is-
land 1
Prairie Is-
land 2
Monticello
Mo.
(None)
Neb.
Cooper
Fort Calhoun
N.D.
(None)
S.D.
(None)
Pre-
1970
1970
1971
1972
330CPWR)
1973 1974
530(BWR)
1975
1976
1977
1978
1979
1980
330(PWR)
S45(BWR)
457(PWR)
778(BWR)
Region Total
543
987
1308 S30
0*
0»
0«
0*
0*
~New Purchases may be expected.
<
I
-------
TABLE IV-7
CENSUS REGION- WEST SOUTH CENTRAL
(Arkansas, Louisiana, Oklahoma. Texas)
State
A: k.
La.
Ok la.
Ten.
Nuclear
Plant Name
Ark. Nuclear!
Ark. Nuclear2
Waterford 3
(None)
(None)
Pre-
1970
1970
1971
1972
1973
820(PWR)
1974
1975
920(PWR)
1976
ues(PwR)
1977
1978
1979
I960
Region Total
820
920
1165*
0*
0»
0*
•New purchases may be expected.
<
I
00
-------
TABLE IV-8
CENSUS REGION: MOUNTAIN
(Arizona, Colorado, Idaho, Montana, Nevada, N. fflTexlco, Utah, Wyoming)
Nuclear
un
Plant Name
Lrlz.
(None)
:oio.
Fort Saint Vraln
da.
(None)
rionl.
(None)
Jev.
(None)
>(. M.
(None)
Jtab
(None)
Myo.
(None)
Pre-
1970
1970
1971
1972
S30(HTCR5)
1973
1974
1975
1975
1877
1978
1979
19B0
Region Total
330
0»
0«
0»
o«
0»
•New purchasei! may be expected.
<
i
CD
-------
TABLE IV-S
CENSUS REGION PACIFIC
(California, Oregon, Washington)
Nuclear Pfe-
aiatc Plant Name 1&70 1970 1971 1972 1973 1974
CjI. Diablo Can- iOSO(PWA)
yon 1
Djablo Can-
yon 2
Ore _
Wash-
ftancho Scco
San On of re 2
San Onofre 3
San Onofre 1
Bay
Point Arena 1
& 3
Trojan
Hanford 2
N Reactor
fi0t{PWR)
(1967)430
-------
NUCLEAR PLANT
CAPACITY
(KILOWATTS)
OPERABLE
9.131,800
BEING BUILT
46,605,000
PLANNED REACTORS ORDERED
48,524,000
TOTAL
104,260.800
total electric utility capacity as of
JULV 31, 1971 3SS.2SB.21B KILOWATTS
OPERABLE ¦ (22)
BEING BUILT A (55)
PLANNED (Reactors Ordered] • (49]
U.S. Atomic Energy Commission
September 30, 1971
Figure IV-1. Nuclear Power Reactors in the United States
-------
IV-12
REFERENCES
IV-1. "Central Station Nuclear Plants - Units Operable, Under
Construction or On Order, " U. S. A. E. C. Division of
Industrial Participation, March 31, 1971.
IV 2. "Nuclear Reactors Built, Being Built, Or Planned in the United
States as of December 31, 1970, " TID 8200 (23rd Rev.), Office
of the Assistant General Manager for Reactors of the U.S. A.E.C.,
1971.
IV-3. Lyerly, R.L., "A Listing of Commercial Nuclear Power Plants,
Edition No. 3, " Southern Nuclear Engineering, Inc., August 1970.
-------
V-l
V. LONG-TERM PROSPECTS
A. Comparison of Nuclear Power Projections
There have been many attempts to project the future requirements for
nuclear power. In this section these various projections are presented for
comparative purposes along with their underlying sources. Several current
estimates of nuclear power requirements were not included since there was
no source or basis given for the quantities. Basic references in which energy
forecasts have been presented and on which others have based their forecasts
are listed as References V-l through V-15.
Along with this wealth of forecast sources,there has been made a
corresponding range of assumptions, and a variety of terminology has been
employed. Thus, confusion arises during the attempt to compare and evaluate
the magnitudes of nuclear power required in the future years. Fundamentally,
there is an electrical energy demand and an electrical energy supply. Electri-
cal energy demand or requirement is a function of population growth and per
capita consumption. Electrical energy supply or availability is a function of
natural resources available, market place considerations, and the development
of technology. In this report an attempt will be made to adhere to this termi-
nology.
Pacific Northwest Laboratories (PNLJ, in Ref. V-l6,collected and compared
the projections of a number of forecasters. They then assessed the methodologies
employed. Most of the forecasts studied (including some from the basic listing
of references above) provided only limited information about their methodology
and practically none provided quantitative statements of the actual forecasting
relationships. Very few forecasts provided standard errors of estimate of
other measures of uncertainty. Some forecasts gave ranges but no information
on the probability that future values would be within the range. Many projections
do not set forth their underlying assumptions nor do they all consider the same
factors in the construction of their projection. Indeed it was apparent that
variations in definitions for such terms as energy consumption, energy re-
quirement, and energy demand existed.
-------
V- 2
The variations in terminology become evident when functional relation-
ships for nuclear power projections were expressed for some of the forecasts
whose abstracts appeared in Ref. V-16. These are listed in Table V-l. In
parentheses is the abbreviation of the reference assigned by the authors of
Ref. V -16. The corresponding page of the FNL report is also given. As an
example, in the EUS reference (Ref. V-13), installed capacity is a function
of the 1960-1985 time period, per capita energy consumption, the population
growth, and future fuel costs.
The first step in the preparation of the curve showing the forecasts of
future nuclear power requirements was to determine the present and near
term installed capacity. Figure V-l shows the growth of this on-line nuclear
capacity. The solid bars show the net nuclear capacity installed in a given
year. These magnitudes were determined from Tables IV-1 through IV-9 by
summing the total capacities shown for all nine census regions for each year.
The cumulative nuclear generating capacity curve was then determined by the
summation of all the installed capacities up through a given year. As antici-
pated, the cumulative capacity curve tails-off in the latter years shown since
more plant purchases can be expected but are unknown at this time.
The cumulative capacities from Figure V-l were plotted on Figure V-2
as the near-term installed capacity that will be available. Then the forecasts
of future nuclear power requirements were plotted from the various sources.
The first 10 sources are identified by the abbreviations assigned by Ref. V-16.
The additional references, V-17 to V-19, discussed in the PNL report were
not previously assigned the designation "basic references" as per this report.
More recent forecasts of installed nuclear power requirements attributed
to Gambs and Rauth, Earhart, and Yarcosaare found inRefs.V-20 through V-23.
In order to satisfy the future demand as shown by the trends of these data, the
nuclear generating capacity available must be an extension of the solid line.
Ideally, this extended curve (now shown) of future nuclear power available would
be an upper bound curve to the required nuclear power data points. Later in
Section V.B, it is shown, after employing a linear programming computer
model, that future nuclear generating capacity at minimized cost can be a
family of curves. The family of curves results from the choices of various
reactor types being developed and employed for future power generation.
-------
V -3
TABLE V-l. NUCLEAR FORECAST FUNCTIONAL
RELATIONSHIPS
Installed capacity
{EUS, p. 57)
F (1960-85 period, per capita con-
sumption, population growth, future
fuel coats)
12
Energy demand (10 B)
(RAF, p. 61)
Energy demand
(EMUS, p. 66)
F (fuel source)
F (1947-65 period, fuel uses, utility
electricity uses, raw material
non-fuel and non-power uses, sector,
sourc e)
Nuclear power capacity
(FGNP, p. 38)
F (present and future installed
capacity)
Gross energy
{PEC, p. 39)
£ all types of commercial energy
Electrical energy require-
ments (NPS, p. 53)
F (population, GNP, geographical areas)
Energy consumption
(PCCP, p. 76)
F (population, labor force, producti-
vity, GNP, Index of Industrial
Production & Gross Product
Originating)
-------
V-4
OJ
£
100,OOOt-
90,000 -
80, 000
70, 000
60,000
u
a
(13
U
lao
c
rH
a)
u
a
c
tu
O
u
ai
4J
(J
D
50, 000
40,000
2 30, 000
20, 000
10, 000
J
P
I
z
H
Ami
2
EB
Pre-1970197019711972 1973 197419751976 IS77 B78 19791980
Year
Key ~ Cumulative Capacity
m Net Capacity Installed
Iin Given Year
* More purchases for these
years can be expected
Figure V-l. Growth of On-Line Nuclear Capacity
-------
V-5
1970 1980 1990 2000 2010 2020 2030 2040
Year
Figure V-2. Forecasts of U.S. Nuclear Power Requirements
-------
V-6
The nuclear energy requirements from the RAF, PEC, CGAEM, and
PCCP references were given in Ref. V-16 in the units of trillions of Btu's.
These are converted to capacities and plotted on Figure V -2 after assuming a
heat rate = 10200 Btu /kwh and an 80 percent load factor <7000 hr/yr plant
operation). For the first 10 references listed on Figure V-2, two assumptions
generally applied are:
(1) Rate of growth of the gross national product (GNP) equal
to ^ 4 percent/year
(2) Projected population growth from the Bureau of Census
equal to "1.6 percent/year
The forecasts for total electricity requirements in the United States are
presented on Figure V-3 for comparison purposes with Figure V-2. The future
total electricity requirement consists of contributions from fossil, nuclear
and hydroelectric powe r utilities, and industrial power generation.
Many future perturbations affect the accuracy of the nuclear power
forecasts. Some of these are:
(1)
Environmental constraints
(2)
Siting
(3)
Nuclear waste disposal
(4)
Politics
(5)
Labor availability
(6)
Inflation
(7)
Fuel availability and cost
They are
such as to cause delays in meeting future power demands. The
are strongly dependent on the commercial marketing assumptions including
those concerning the establishment of large fuel fabrication and reprocessing
industries, acceptance of safety and siting features, and the availability of
trained personnel and resources. Timely introduction of the breeder reactor
is necessary to reduce power costs by substantial margins while providing
good nuclear fuel utilization. Environmental factors now play a significant
part in power plant selection as well as costs. In a case such as California,
-------
V-7
Figure V-3, Forecasts of U. S. Total
Electricity Requirements
-------
V-8
such stringent smoke control ordinances have been passed that no more fossil
fuel plants are possible without major control equipment. In other states, such
as Maryland and southern Florida, some groups are deeply concerned about
the thermal effects of nuclear power plants on surrounding bodies of water.
Also, in the future it is possible that a constant or even an increasing amount
of energy per unit GNP may be required if present practices of encouraging
the use of energy are continued.
B. Effects of Reactor Types and Their Development
on Energy Projections and Costs
A linear programming (LP) model of the U.S. electrical power economy
has been developed under the Civilian Nuclear Power (CNP) program. The
model was applied (Ref. V-24) to evaluate the benefits of alternate courses of
electrical power system developments. The reactor types treated in the study
were:
(1) Light water reactors (LWR)
(2) Heavy water reactors (HWR)
(3) High temperature gas-cooled reactors (HTGR)
(4) Liquid metal fast breeder reactors (LMFBR)
(5) Gas-cooled fast breeder reactors (GCFR)
(6) Steam cooled fast breeder reactors (SCFR)
(7) Molten salt breeder reactors
A 50-year period, from 1970 to 2020, was projected. Optimum cost
cases are presented and the nuclear percentage of future total power capacity
(installed after 1970) was estimated for each case.
Some of the ground rules and assumptions utilized in the analysis were:
(1) Nuclear power plant designs are based on technology feasible
in 1967.
(2) The linear programming model obtains an optimal solution
over the 1970-2020 period on the basis of minimal costs
discounted at seven percent.
-------
V-9
(3) Price inflation or escalation is not accounted for.
(4) LWR's are represented by PWR's.
(5) The rising uranium price schedule assumed is based on
information from the A. E.C. 's Division of Raw Materials.
(6) Power plants built before 1970 will not be included in the
study, but the effect of their omission is reflected in the
capacity factor.
(7) LP model determines best load factor history for each power
plant so that the present worth of cash flows will be minimized.
(8) LP model selects plants to minimize total discounted power
cost over study period.
(9) UgOg cost will be represented as a function of cumulative
consumption and predicted reserves.
(10) Costs are based on June 1967 prices and privately owned
utility practices.
(11) Nominal plant size will be 1000 Mwe.
(12) Power plant life will be 30 years.
The ground rules, selected in 1967, assumed a capital cost of $133/Kw
for a 1000 Mwe pressurized water powe r plant and $100/Kw for a 1000 Mwe
coal-fired plant. In the latter part of 1970 for the 1000 Mwe size range, the
capital cost of a nuclear plant is about $250/Kw and a coal-fired plant, $195/Kw.
While nuclear fuel costs have remained essentially constant, fossil fuel costs
have increased dramatically and are now much higher than those used in the
study. Since inflation applies to the entire electric power industry, it has
been hypothesized in Ref. V-24 that inflation would not have a marked effect
on the power plant mix developed in the study. Also in reviewing the cost
estimates in the study, one should consider the relative cost differences from
case to case and not their absolute magnitudes.
The major part of the LP model input data includes prediction of plant
performance, capital costs, fuel cycle costs, and introductory dates for new
reactor types. It was provided by the assigned task forces under the CNP
-------
V-10
program that were charged with the evaluation of light water reactors,
advanced converter reactors, the liquid metal fast breeder reactor, alternate
coolants to sodium for fast reactors, the role of thorium, and fuel recycle.
The task forces generated this information on the basis of specific 1000 MWe
reactor designs which were evaluated for each of the reactor concepts con-
sidered in the studies. These designs were based on information provided
by proponents of the systems and therefore generally reflected their view-
point and enthusiasm. The evaluations of the reactor types are presented
in Refs. III-2 through III-4 and III - 6.
The computer analysis of Hef. V-24 treated three categories of cases:
(1) Fossil only
(2) Combined nuclear and fossil
(3) Nuclear only
The following discussion is only concerned with results from the first two
categories, and the reader is referred to WASH 1098 for the complete details
of these detailed analyses. The case 1-A with only fossil fueled power
plants was necessary to establish a reference power cost in a system without
nuclear power. Performance in the model is based upon average capital costs
of $100/Kw for coal-fired plants and $90/Kw for gas-fired plants and upon fuel
costs ranging from 1.6 to 3.4 mills/kwh depending upon the geographical
region. Two variations in the cost of power supplied by fossil fuel were also
examined:
(1) Perturbation due to a 1 percent per year increase in fuel
cost
(2) Perturbation due to an increase in capital costs by
$10/Kw
Thus, the sensitivity of the reference fossil-fuel case to fuel and capital costs
was obtained.
Seven cases, 2-A through 2-G, with both fossil and nuclear power plants
available were computed. Thus, an assessment of the major possible devel-
opments in reactor technology in competition with fossil was obtained. The
cases combining nuclear and fossil power plants are defined at the top of Table
V-2.
-------
TABLE V-2. RESULTS OF SYSTEMS ANALYSIS TASK FORCE CALCULATIONS FOR
FOSSIL AND NUCLEAR PLANTS WITH RISING URANIUM COSTS
'ase Number
Case Definition
Plants Included
1 -A
Fossil
2-A
Fossil
LWR
2-B
Fossil
LWR
LMFBR
(Reference
Oxide only)
2-C
Fossil
LWR
LMFBR
2-D
Fossil
lwr
LMFBR
HTGR
HWOCR
2-E
Fossil
LWR
LMFBR
HTGR
HWOCR
GCFR
SCFR
MSR
2-F
Fossil
LWR
HTGR
LMFBR
2-G
F ossil
LWR
LMFBR
HTGR
Characteristics of Optimum Solution
Total Cost. 1970 Present Worth, billions
1970 to 2020 at 7% Discount Rate
Levelized Power Cost
1970 to 2020 mills/Kwh
Total Nuclear Capacity, Mwe
1355
2000
% of Total Capacity in 2000
Fossil
LWR
HWR
HTGR
LMFBR
GCFR
SCFR
MSR
Total Uranium Used, Thousands of Tons U„0„
Through 2020 a-B-
Max. Price through 2020, $/lb UjOg
Ave. Price through 2020
215.3
4.939
100
205. 3
4. 688
I24xl03
484xl03
64. 2
35. 8
1906
37.50
16.73
201.4
4.599
.3
44.4
22.2
33.2
1568
27.50
12. 00
189.0
4.316
3
155x10 158x10
760x103 1.056xl03
21.9
20.9
57.2
979
13.75
11.34
185.6
4.238
I58xl03
1.O59X103
21.6
7.8
7.4
20.7
42.5
1255
22.50
14.04
178. 9
4.085
162xl03
1.122X103
17.0
8. 2
8. 0
5. 9
32. 8
2.4
25. 7
1004
17.50
11. 70
179. 8
4.106
171*10,
1. 155xl03
14.6
10. 7
25. 1
49.6
1199
17. 50
107.7
4.286
12. 99
143x10^
1, 062x10
22.7
8. 4
44. 4
24.5
1616
32.50
16. 32
<
I
-------
V-12
Case 2-B considers only the reference oxide design of the LMFBR which
would be available in 1980, thus limiting LMFBR development to an early
design. Case 2-C allows for the full development of the LMFBR technology.
Thus for this case the following sequence of plant introductions is considered:
(1) The reference oxide LMFBR in 1980
(2) The reference carbide design in 1984
(3) The advanced oxide design in 1990, and
(4) The advanced carbide design in 1994
Case 2-D considers the introduction of the advanced converters in 1976
which then displace the LWR to the extent allowed by new reactor growth.
Case 2-E treats fossil plants and all of the reactor types. All dases up to
this point consider that for each specific reactor type, the fuel costs are
subject to both rising uranium costs and improving fuel technology, but the
capital costs remain essentially fixed. Cases 2-F and 2-G, in addition to the
uranium and fuel technology assumptions, include the effects of improving
technology, experience, and unit size on capital costs. Both cases, therefore,
use declining capital costs for the LWR's and HTGR's. Also, LMFBR capital
costs are assumed to be constant at $135/Kwe and $165/Kwe, respectively,
for Cases 2-F and 2-G.
The results of the systems analysis employing the LP model are
summarized in Table V-2. The most meaningful cost evaluation parameter
shown in the table is the discounted cost of power. This is defined as the
present worth of all power costs between 1970 and 2020. The levelized power
cost in mills /Kw-hr is defined as the total present worth of all power costs
divided by the present worthed energy production. Since the present worth
of all energy is a constant for all cases studied, any solution which yields
the minimum discounted cost of power must also give the minimum levelized
power cost. For each case considered, the selection of reactor types was
such that the total power costs would be minimized over the period 1970-2020.
It is seen from Table V-2 that the minimum cost case is 2-E, an optimum
combination of fossil plants and all types of reactors.
The total or cumulative nuclear capacity (Mwe) is given in Table V-2
for the years 1980 and 2000. Cumulative nuclear capacity for each case
-------
V-13
is also plotted in Figure V-4. The cumulative LWR sales through 1972
are incorporated into the model and consequently into all cases except 1-A:
12,000 Mwe by 1970 and 34, 000 Mwebyl972. These magnitudes are somewhat
higher than the values read from the near-term installed capacity line of
Figure V-l. The curve for Case 2-A results as shown because the LWR
captures the market for heavy load power in every power supply area during
the 1970's and 1980's. However, with a rise in uranium prices, the fossil
plants recapture the market. Thus, LWR's supply the demand only in the
high cost fossil areas and only through the mid 1980's. Table V-2 also
presents for each case the uranium quantity used and U^Og unit costs.
The nuclear percentage of the future total power capacity installed after
1970 is shown in Figure V-5. This percentage is itemized per .reactor type
and is shown for each case for the year 2000 in Table V-2. In the trivial
case 1-A, this percentage is zero. Figure V-5 indicates that for Case 2-A
the LWR's supply 45 percent of the total generating capacity (installed after
1970) by the year 1980, approximately 37 percent by the year 2000, and
15 percent by the year 2020.
In Case 2-A, where only the LWR is available, increasing uranium
prices force the LWR out of the optimum generating system. However, in
Case 2-B the reference oxide LMFBR, by holding down the uranium con-
sumption, makes possible an increased penetration of nuclear plants into the
generating system (Figure V-4). Also after the year 2002 the reference oxide
LMFBR produces plutonium for new reactors at a rate greater than that re-
quired by the economy. Thus, it will be economically attractive to burn
the excess Pu in plutonium-fueled light water reactors. Consequently,
it is calculated that about 100 reactors fueled by excess plutonium from the
reference oxide LMFBR will be constructed by the year 2020.
In Case 2-C, the timely introduction of each successively, advanced
LMFBR design causes not only that design to replace all previous LMFBR
designs, but also breeder reactors to capture more of the total capacity
market. Thus, by the year 2004, no more fossil plants are built for the
nation's power economy. As in Case 2-B, overproduction of plutonium
breeders causes plutonium-fueled light water reactors to be introduced after
the year 2002.
-------
V -14
3200 -
2800 -
2400 -
2000 -
1600 -
1200 -
800 -
400 --
1970
1980
2010
2020
Case 1A
Case 2A
Case 2B
Case 2C
1990 2000
Year
Fossil Case 2D - Fossil + lwr + lmfbr+ advanced conv.
Fossil + lwr Case 2E - Fossil + lwr + hwr + htgr f gcfr -taisr +¦ scfr
Fossil + lwr + lmfbr(RO) Case 2F - Fossil + lwr + htgr + lmfbr($135/KW)
Fossil + lwr + lmfbr (ALL) Case 2G - Fossil + lwr + htgr + lmfbr($165/KW)
Figure V-4. Cumulative Nuclear Generating Capacity
-------
V -15
10Q
1970
1980
1990
2000
2010
2020
Case 1A - Fossil Case 2D - Fossil + lwr + lmfbr + advanced conv.
Case 2A - Fossil + lwr Case 2E - Fossil + lwr + hwr + htgr + gcfr + msr +
Case 2B - Fossil + lwr 4-lmfbr (RO) Case 2F - Fossil + lwr + htgr + lmfbr <$135/KW)
Case 2C - Fossil + lwr + lmfbr (ALL) Case 2G - Fossil + lwr + htgr + lmfbr ($165/KW)
Figure V-5. Nuclear Percentage of Future Total
Power Capacity Installed After 1970
-------
V-16
In Case 2-D advanced converters are represented by the HTGR, the
heavy water-moderated organic-cooled reactors (HWOCR) and the heavy
water-moderated boiling-light water-cooled reactor (HWBLW) designs.
They are introduced in 1976 and displace the LWR to the extent allowed by
new reactor growth. The advanced converters displace most of the reference-
oxide LMFBR's and some advanced-oxide LMFBR's. This requires more
uranium than that saved by displacing the LWR's, so the uranium consumption
by the year 2020 increases. When the advanced-carbide LMFBR becomes
available, the advanced converter construction stops. As in Case 2-C,
fossil-fueled plants are no longer built after the year 2004 because they
are competing against the same advanced carbide LMFBR. The system
begins to build Pu-fueled LWR's by the year 2002. After this time, nuclear
plants in Case 2-D capture about the same percentage of the total installed
market as in Case 2-C (Figure V-5).
Three additional designs (GCFR, MSR, andSCFR) were made available
to the power economy in Case 2-E. The Pu-fueled molten salt converter (MSR)
and the advanced gas cooled fast breeder (GCFR) were built in large numbers,
while only a few steam-cooled breeders were built (SCFR). The advanced
GCFR has a doubling time equal to that of the advanced LMFBR and an
estimated $20/Kwlower capital cost. Because of this lower capital cost,
the GCFR is essentially the only breeder built in Case 2-E. The excess
plutonium produced by the GCFR is consumed by the Pu-fueled MSR which
thus displaces the LWR. In Case 2-E both advanced converters (HTGR and
HWR) are made available but to a lesser extent than in Case 2-D. They
are eventually displaced by the MSR.
Cases 2-F and 2-G were designed to assess how the LMFBR would
offset power costs when the LWR and HTGR plant costs are decreasing with
time. In Case 2 -F the capital cost of the LMFBR is assumed to be $135/Kw,
that of the HTGR to drop from $122 to $102 per Kw, and that of the LWR to
drop from $130 to $104 per Kw. In this case, the HTGR's are built at the
maximum rate, and the LWR's are initially built in great numbers so that
in 1980 the total nuclear capacity reaches 171, 000 Mw (Figure V-4). No LWR's
are built after the 1982-83 period, and thereafter the HTGR and the LMFBR
share the nuclear market until the year 2000. The HTGR has lower power
-------
V-17
costs than either the reference oxide or reference carbide LMFBR's, but
these are built at the maximum allowable rate in order to function as
plutonium producers. However, since the building rate of LMFBR's is
limited by plutonium availability, the number of HTGR's being built in the
1980's is large. Even in the early 1990's, when the advanced oxide LMFBR
becomes available, the HTGR's are still being built. It is worthwhile to note
that no LWR's are built to provide plutonium in this period, even though their
capital costs are approximately the same as those of the HTGR's. The reason
for this is that the rising uranium costs ($111 lb by 1988 and $17. 50/lb by
2000) make the LWR's too costly. Although the advanced carbide LMFBR
essentially supplies all new capacity after 1994, some HTGR's are still
being built in the 1996-97 period becuase of plutonium limitations. However,
these temporary plutonium shortages do not prevent a plutonium surplus
in the nuclear system around the year 2000, and LWR's are built again as
plutonium burners. The discounted power cost to year 2020 is $179. 8 billion.
This is $5.8 billion less than in Case 2-D. The nuclear share of the power
economy market is about 3 percent more than in Case 2-D by 1980, about
8 percent more by 2000, and about 3 percent more by 2020.
Increasing the capital cost of the LMFBR to 165/Kw in Case 2-G makes
it advantageous to build primarily HTGR's from 1982 to 1995 because the
HTGR's have an assumed capital cost of about $50/Kw less than that of the
LMFBR's. Although some reference carbide and advanced oxide LMFBR's
are also built as plutonium producers, HTGR's are built in large numbers
as late as year 2004 because of plutonium limitations. The advanced carbide
LMFBR design, which would capture all the nuclear market after its intro-
duction, is held back by a shortage of plutonium. The system in Case 2 -G
does not obtain a surplus of plutonium until 2008, at which time LWR's are
built again as plutonium burners. What makes nuclear capacity in year 1980
3
drop to 143x10 Mwe is increased uranium prices. This uranium price increase
is due to increased uranium requirements brought about by large numbers of
LMFBR's being displaced by HTGR's, and so making it uneconomical to build
more LWR's. Nuclear power plants in Case 2-G capture about the same market
by year 2000 as in Case 2-D and then lose about 4 percent of it by year 2020.
The discounted power cost in Case 2-G increases to $187.7 billion, which is
$7. 9 billion more than in Case 2-F.
-------
V -18
The advanced carbide LMFBR, even at a capital cost of $165/Kw,
supplies the required generating capacity upon its introduction. However,
the incentive for its rapid buildup is much greater at $135/Kw than at
$165/Kw. Case 2-F reflects this in the increased number both of oxide
and carbide LMFBR's and of plutonium-supplying LWR's and in the de-
creased number of HTGR's.
Cases 2-F and 2-G also indicate that a high-performance breeder such
as the advanced carbide LMFBR with its low inventory, short doubling time,
and insensitivity toward changing fissile material costs can:
(1) Probably capture a major portion of the market even with
capital costs that are about $65/Kw greater than those of
fossil-fueled plants and about $40/Kw greater than those of
the HTGR or LWR
(2) Rapidly approach and meet the total electrical demand, even
in the absence of a large number of plutonium producing
reactors that supply plutonium for its growth
In summary then, Cases 2-A through 2-G demonstrated that fossil-
fueled power plants and LWR nuclear plants only can produce moderate
total savings. In order for nuclear plants to capture and hold a major portion
of the electrical power market and provide discounted savings of $30-40 billion
by the year 2020, high-performance breeders will be required. It appears
(Figure V-5) that these high-performance breeders will permit nuclear plants
to capture over 70 percent of the market (installed capacity after 1970) by
year 2000 and over 90 percent by year 2020, assuming that fossil-fuel prices
remain constant and that capital cost differentials between nuclear and fossil-
fueled plants range between $20 and $50/Kw. Advanced converters offer
savings when capital costs are constant and the uranium prices are rising.
The nuclear power forecasts shown in Figure V-5 show the strong
dependency upon reactor type technology and development. These can be
compared to the forecasts in Figure V-2. In the cases treating nuclear power
only, it was shown that a six-year delay in the anticipated schedule of fast
breeder development results in a substantial decrease in the savings associated
with the breeder.
-------
V-19
C. Fuel Availability
1. Uranium
Estimates of uranium resources available as of January 1, 1970 are
shown in Table V-3 (Ref. V-24). Reasonably assured resources are in known
ore deposits and occur in such grade, quantity, and configuration that they can
be profitably produced with current technology at the given prices. At $8/lb
and $10/lb prices, reasonably assured resources can be considered equivalent
to ore reserves in the usual sense. Data on resources in the higher price
ranges are not as well developed, and the estimates are less accurate.
Estimated additional resources are in the extensions of known deposits
or in undiscovered deposits in known or postulated uranium districts. Es-
timated resources in the price category of $15/lb and lower are almost
entirely tabular ore bodies in sedimentary rocks of the Western States.
By-product resources include uranium recoverable in conjunction with
production of phosphoric acid, principally in Florida, and uranium recoverable
from copper leach solutions. Projected production from such by-product
sources, estimated to be available through the year 2000, is included in the
estimates. Some of this by-product is expected to be available at prices in
the $8-$10/lb range.
Research work indicates that uranium recovery from the oceans (about
4000 million tons of U^Og) may be possible at the high price range. Although
there are no foreseen needs to exploit these resources, they do represent a
virtually inexhaustible source of high-cost fuel which eventually could be
used in breeder reactors if lower cost resources become exhausted.
The quantities of U^Og fuel required each year were projected by
Gambs and Routh (Ref. V-21). These are presented in Table V-4:
-------
TABLE V-3. ESTIMATED U. S. URANIUM RESOURCES
UgOg Price
Per Pound
$ 8.00
10. 00
15. 00
30. 00
50. 00
100.00
January 1, 1970
Cumulative Thousands of Tons of U^Og
Reasonably Assured
Conventional
Deposits
204
250
3 90
530
5, 400
11, 400
By-
product
90
110
110
110
110
Total
204
340
500
640
6, 000
12, 000
Estimated Additional
Conventional
Deposits
390
600
950
1, 600
4, 000
13, 000
By-
product
Total
390
600
950
1, 600
4, 000
13, 000
Total
594
940
L, 450
2, 240
10, 000
25, 000
<
I
to
o
-------
V -21
TABLE V-4. QUANTITIES OF UgOg FUEL
REQUIRED EACH YEAR
Year
UgOg Quantity, tons/yr
1965
1970
1975
1980
1985
1990
1995
2000
30,000
65,000
100, 000
140,000
150,000
200
600
9,000
The cumulative quantity of UgOg fuel required up to the year 2000 was
determined from these data to be 2100 x 10^ tons. Thus from Table V-3
one sees that the uranium resources are abundant enough in the U.S. such
that it is not necessary to consider any material above a cost of $50/lb.
2. Thorium
Although there has been little demand for thorium to date, and hence
little prospecting for it, the estimated U.S. thorium resources are large, as
shown in Table V-5. They are located predominantly in Lemhi County, Idaho,
and in neighboring Montana, where preliminary investigation indicates the
presence of deposits containing some 100, 000 tons of ThOg from which thorium
can be produced at prices comparable to present-day prices for uranium.
TABLE V-5. ESTIMATED U.S. THORIUM RESOURCES
Millions of Short Tons
$/lb ThOg
10
30
50
100
Reasonably
Assured
Estimated
Additional
Total
0.1
0. 2
3.2
11.2
0.5
O.fi
7. 6
24.6
0. 6
0. 8
10.8
35. 8
-------
V -22
3. Bred Fuels
The bred fuels, plutonium and U-233, are produced by certain reactor
types and are consumed by others. The LMFBR and GCFR reactors use
uranium-238 fuel and breed plutonium-239 which can then be consumed in
Pu-fueled LWR's. Thermal breeder reactors use thorium-232 fuel and
breed uranium-233.
-------
V -23
REFERENCES
Basic References on Forecasts of
Nuclear Power Requirements
V-l. "Statistical Year Book of the Electric Utility Industry for
1969, " Edison Electric Institute, New York, September 1970.
V-2. "National Power Survey, " The Federal Power Commission,
Washington, D.C., 1964, 1969, and 1970. (NPS)
V-3. "The Economy, Energy and the Environment, 11 Joint Economic
Committee, Congress of the U.S., September 1, 1970.
V-4. Landsberg, H. H., et. al., "Resources in America's Future,
Patterns of Requirements and Availabilities 1960-2000,11
The Johns Hopkins Press, Baltimore, 1963. (RAF)
V-5. "Nuclear Power and the Environment,11 U.S. Atomic Energy
Commission, Division of Technical Information, 1969, p. 6.
V-6. Roddis, L.H., "The Future of Nuclear Power, " Science and
Technology Advisory Council to the Mayor, New York, New York.
V-7. "Considerations Affecting Steam Power Plant Site Selection, "
Office of Science and Technology, December 1968.
V-8. "Research and Development Problems in Power Plant Siting, "
Office of Science and Technology, Washington, D.C., 1970
(to be published).
V-9. Civilian Nuclear Power, The 1967 Supplement to the 1962
Report to the President, February 1967, U.S.A.E.C. (CNP)
V-10. Sporn, P., "Nuclear Power Economics -Analysis and Comments-
1967, " pp. 2 20, Joint Committee on Atomic Energy (U.S.)
Report Nuclear Power Economics-1962 thru 1968, February 1968.
V-ll. "Cost Benefit Analysis of the U.S. Breeder Reactor Program, "
WASH 1126, U.S.A.E.C. Division of Reactor Development and
Technology, April 1969.
-------
V-24
REFERENCES
Basic References on Forecasts of
Nuclear Power Requirements (Continued)
V-12. "An Energy Model for the U.S., " Bureau of Mines, U.S. Depart-
ment of Interior, Report IC 8384, July 1968. (EMUS)
V-13. Michael C. Cook, "Energy in the U. S., 1960-1985, " Sartorxus
& Company, September 1967. (EUS)
V-14. "Projection of the Consumption of Commodities Producible
on Public Lands of the U.S., 1980-2000, " Robert J. Nathan
Assoc., May 1968. (PCCP)
V-15. "Forecast of Growth of Nuclear Power, " WASH 1084, U.S.A.E.C.
Division of Reactor Development and Technology, December 1967.
(FGNP)
Additional and More Recent References
Giving Nuclear Power Forecasts
V-16. "A Review and Comparison of Selected United States Energy
Forecasts, " Prepared for the Office of Science and Technology
by Pacific Northwest Laboratories of the Battelle Memorial
Institute, December 1969.
V-17. "United States Petroleum Through 1980," United States Depart-
ment of Interior, Office of Oil and Gas, July 1968. (USP)
V-18. Vogely, W. A., "Patterns of Energy Consumption in the U. S., "
United States Department of Interior, Division of Economic
Analysis of the Bureau of Mines, 1962. (PEC)
V-19. "Competition and Growth in American Energy Markets, 1947-
1985, " Texas Eastern Transmission Corporation, 1968. (CGAEM)
V-20. Gambs, G.C., "The Electric Utility Industry: Future Fuel
Requirements 1970-1990," Mechanical Engineering, April 1970,
pp. 42-48.
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V -25
REFERENCES
Additional and More Recent References
Giving Nuclear Power Forecasts (Continued)
V-21. Gambs, G.C. and A. A. Routh, "The Energy Crisis, 11 Chemical
Engineering, May 31, 1971, pp. 55-68.
V-22. Earhart, J. P., "Basis for Projected Air Pollution from Com-
bustion of Fossil Fuels in the U.S.,11 Air Pollution Control
Office of the Environmental Protection Agency, June 1970.
V-23. Yarosh, M. M., "Changing Emphasis in the Siting of Steam
Electric Power Plants, " A. S. M, E. Paper No. 70-WA/Ener-12,
November 1970.
V-24. "Potential Nuclear Power Growth Patterns, " WASH 1098, U.S.
A. E.C. Division of Reactor Development and Technology,
December 1970.
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VI-1
VI. SAFETY CONSIDERATIONS, SPECIAL
LICENSING REQUIREMENTS. DELAYS
Due to the potential hazards of radiation, the safeguards which are engi-
neered into nuclear plant systems are multiple redundant, fail-safe, and sub-
ject to rigorous exhaustive and frequent testing, making the operation of a
nuclear power plant among the safest and most reliable of power generating
systems in terms of safety considerations. During the preliminary and final
licensing procedures of nuclear power plants, these safety systems are re-
viewed several times for assurance of operating safety. However, because
of the engineering design and lengthy evaluation of these systems, a signifi-
cantly larger amount of engineering time is required to plan such systems
and, because of the multiplicity of safety measures, often interlocked, sub-
stantially more time is required to construct and test a nuclear plant prior to
placing it in service.
A nuclear-fueled steam electric plant, because of the nature of its fuel
and the by-products of the fission reaction, comes under the licensing control
and regulatory authority of the Atomic Energy Commission (Ref. VI-1). In
order to protect the public and the environment, these regulations and opera-
ting guidelines are rigorously enforced and observed.
Fossil-fueled steam electric plants release to the environment the products
of combustion, including particulates, carbon monoxide, carbon dioxide, nitrogen,
and sulfur oxides, and others. A nuclear-fueled plant releases none of the above
contaminants to the environment, but does release very minor amounts, care-
fully monitored and regulated, of the products of fission, including tritium,
iodine-131, krypton-85, xenon-133, and other noble gases and their daughter
products. The need to monitor, regulate, and control these radioactive
materials within the nuclear plant and upon release to the environment predi-
cates a highly sophisticated waste-product handling and monitoring system
unlike any safety system found in fossil-fueled plants and invariably results
in substantially greater initial capital investment and in greater construction
complexities and time periods.
Presently, the construction of a large fossil generating unit may take
from four to five and one-half years, depending on several factors such as the
need for additional capacity, the region of the country, and the fuel to be
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VI-2
burned. The construction of a nuclear power plant may take as long as seven
years to construct, partially due to stringent licensing procedures and com-
plexity of construction, but also attributable in several instances to the inter-
vention of groups or individuals intent on the fullest measure of protection to
the environment. Because of the predictable delays due to various causes and
because of the increasing frequency of hearing and review delays through
interventions, the placing of a nuclear unit on-line may take up to seven years
or more. Further, because of the growing backlog of review necessary to
license a nuclear plant, and because of the Atomic Energy Commission's new
responsibility in the enforcement of the National Environmental Protection
Act, it can be anticipated that even longer periods of time will be required in
the future to place a nuclear plant in service. Utilities have recently adopted
the practice of including a delay time in their planning schedule and, in several
recent cases, have allowed as much as 10 years for the construction and
licensing of a nuclear unit.
REFERENCE
VI-1. Section 10, Code of Federal Regulations, Parts 10, 20, 50.
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vn-i
VII. ADVANTAGES AND DISADVANTAGES
OF NUCLEAR POWER
The advantages and disadvantages of nuclear power generation versus
other types of generation are presented. Then the advantages and dis-
advantages of the various nuclear reactor types are discussed. The types of
nuclear power reactors were listed in Table III-l and described in Section III.
In Section V, it was shown that the forecast of power supplied by
nuclear reactors in the future will be an increasing function if nuclear reactor
technology is allowed to develop in an optimum manner. Thus a major
advantage is that adequate sources of uranium and thorium will be utilized as
fuel. The demand on the fossil fuel resources of coal, oil, and gas would be
significantly alleviated. Ideally, by the year 2000 fossil fuel plants could
possibly supply under 20 percent of the cumulative electric generating
capacity installed after 1970. In addition, the class of breeder reactors would
be breeding plutonium fuel for use in other nuclear power reactors.
Fewer fossil-fueled plants installed in the future would also reduce the
air pollutants emitted: sulfur, nitrogen oxides, and ash. Nuclear power
reactors are not contributors to the air pollution problem. However, there
are the problems of nuclear waste disposal and thermal pollution.
The relative merits between the types of nuclear reactors are presented
by considering each class in a separate section. The advantages and dis-
advantages of the particular class is considered with respect to any other type.
A. Light Water Reactors
The competitive marketability of LWR's depends strongly on the continued
availability of uranium fuel at low cost. If the cost of uranium rises in future
years, LWR's will be in a disadvantageous situation.
The loss of coolant accident resulting from a primary system rupture
has been, and continues to be, a subject for accident analysis in LWR licensing
and is a design basis for emergency systems to maintain core cooling.
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vn-2
B. Advanced Converters
As a class, these reactors have two main potential advantages over
present LWR designs in that:
(1) Less uranium ore is required to produce the same amount
of energy
<2) Estimated power costs are lower
Most of the advanced converter concepts have some close similarity
with the light water concept in their reactor and fuel designs and in fuel
processing requirements. However, the HTGR concept is an entirely different
reactor and fuel concept that will require an almost entirely new and separate
base of supporting industry (fabrication, reprocessing, and component manu-
facture) to sustain it. Some recognition of the difficulty of establishing the
HTGR as a viable concept because of this unique problem would be appropriate.
However, the HTGR appears to offer the most cost and fuel resource incentives.
It also has the earliest date of commercial application for the advanced con-
verters.
Heavy water availability, price, and risk of a major heavy water loss
during reactor operation are major factors affecting the marketability of
heavy water reactors. With present known and developed processes, the
capital and operating costs of heavy water production plants can be predicted
fairly well. Large plants operating at capacity for their useful life would be
required to realize the assumed low DgO price of $17. 50/lb.
The major DgO requirement for a heavy water reactor is in supplying
the initial inventory. A high DgO production rate and low cost thus implies
a high rate of construction of new heavy water reactors. This results in a
situation in which large D20 plants cannot be justified without sound evidence
that a high rate of heavy water reactor construction will prevail, and such
construction commitments cannot be justified without firm low price commit-
ments for DgO production, implying large-capacity DgO plants. Thus, a kind
of mutually exclusive closed cycle is established and, because of this, the
assumed cost of $17. 50/lb for heavy water is thought to be optimistic.
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vn-3
The risk of a large heavy water loss and the insurance cost to cover
the risk is another factor that appears to have a strong deterrent effect
today on acceptance of the heavy water reactor concepts. Insurance costs
as high as 20 percent per year have been mentioned and some recognition of
this difficulty would seem appropriate.
The enriched uranium fueled HWOCR has economic advantage over
the natural uranium fueled one. The use of thorium in the HWOCR has less
economic potential than for the other advanced converters and LWR's,
The HWOCR designs are quite sensitive to coolant temperature. For example,
a decrease in outlet temperature of 25° F results in a 0. 2 mill/Kw-hr increase
in the energy cost. This illustrates how sensitive the designs are to engineering
judgments and points out the difficulty of comparing such reactor plants.
Another instance of design sensitivity is illustrated by the mixer-strippers
for the HWBLWR plant. Without these mixer-strippers, which yet remain
to be demonstrated, the reactor is economically unattractive.
C. Thorium Systems
The transition from the relatively crude MSRE to a much more complex
full-scale breeder reactor requires an extensive R&D program including
scaleup of components. The MSBR pump design flow rates and power density
would be considerably greater than those in the MSRE. While individual facets
of the technology may have been investigated in the MSRE as well as other
reactor systems, e. g., HFIR, EBR-II, and Dounreay Fast Reactor, it is
only by integrating all the various components and systems in an adequately
sized reactor experiment under conditions similar to those existing in the
actual breeder that the true operating characteristics and potential of the
molten-salt reactors will be determined. To achieve this it would be
necessary to construct a power-producing reactor which would furnish data
on fuel processing, breeding ratio, and secondary coolant behavior that must
be known before the MSBR can be built commercially with confidence. At the
present time the single fluid breeder concept is in the very early design stage.
Thus development of a finalized detailed design of the concept is necessary
before a thorough evaluation can be completed.
While there are no indications that dynamic instabilities will occur,
the dynamic behavior of the system is very complicated, and further accurate
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vn-4
and detailed analysis and experimental work are needed for designing a self-
regulating system that is stable for constant power and also for transient
and load-following behavior.
Pumps and heat exchangers appear to be critical components. While
the MSRE and experimental salt pumps have successfully logged thousands
of hours of molten-salt operation and the MSRE heat exchangers have also
operated successfully for thousands of hours, scale-up to MSBR size and
modifications in design required for the MSBR operating conditions will have
to be demonstrated. The presence of radioactivity, the need for adequate
pressure relief against high-pressure steam, and salt cleanup problems in
case of tube leakage appear to be some of the design and maintenance compli-
cations.
Remote maintenance of a molten-salt fluid-fuel reactor is required due
to the presence of intense gamma radiation in the equipment outside the
reactor caused by activation of sodium and fluorine in the salt, the presence
of fission products, and activation of the structural material by delayed
neutrons in the circulating salt. Pumps and heat exchangers will have to be
capable of long maintenance-free life, as no practical reactor system could
tolerate too many shutdowns due to failure of large components.
D. Breeder Reactors
A big advantage of the breeder reactor is that it produces fissionable
material to refuel itself and in addition to fuel another reactor. The economic
potential of fast breeder reactors lies mainly, but not entirely, m the fact
that they would conserve resources of nucLear fuel. Also,because fast
breeder reactors will operate at far higher temperatures than are encountered
in contemporary water reactors, they will have greater thermodynamic efficiency.
Today's light water reactors operate at an overall efficiency of about 32 percent.
Modern fossil-fueled plants operate at about 39 percent efficiency. Hence,
light water reactors add more waste heat to the environment per unit of
electrical energy produced than fossil-fueled plants do. Fast breeder reactors
will probably attain efficiences equal to that of the most modern fossil-fueled
plant, thereby reducing the nuclear waste-heat problem.
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VII-5
In the LMFBR, the sodium coolant has excellent heat transfer characteris-
tics. Moreover, it can be used at a fairly low pressure even though it emerges
from the reactor at a temperature (about 500 degrees Celsius) that with water
would give rise to high pressures. Indeed, the sodium pressure arises solely
from the force required to maintain the high rate of flow through the maze of
tubes in the core and the blanket. Compared with coolants such as water and
gas, sodium requires low pumping power. It is not particularly corrosive to
the reactor. Because of the inherently low pressure of the sodium coolant,
the reactor vessel and its associated piping need be designed to withstand
only moderate operating stresses, in marked distinction to the pressure vessels
and other primary-system components of a pressurized-water reactor, a
boiling-water reactor or a gas-cooled fast reactor.
Sodium does have certain disadvantages that markedly influence the design
of a reactor. Since sodium is opaque, provision must be made for the mainte-
nance and refueling of the reactor without benefit of visual observation. Sodium
is,of course,highly reactive chemically, and it becomes intensely radioactive
when it is exposed to neutrons, even though its "cross section, " or neutron-
absorption capacity, is relatively low. Hence,the sodium must be kept out of
contact with air or water, and radiation shielding must be used to protect
workers who are near sodium that has been through the core and blanket of
an operating reactor.
Fast breeder reactors cooled by gas or steam are technically feasible
in that no research and development breakthroughs are required. However,
GCFR performance is strongly dependent upon the behavior of the fuel cladding.
Thus,a demonstration of the satisfactory operation of some type of pressure-
equalized cladding is necessary prior to any more comprehensive effort to
develop the GCFR.
Both steam- and gas-cooled reactors are capable of operating with breed-
ing ratios of interest. The gas-cooled breeder reactor, due to its harder
neutron spectrum, can achieve the highest breeding ratio. Breeding ratios of
low pressure steam-cooled reactors can be attractive, if somewhat lower.
The conversion ratios attainable in supercritical pressure steam-cooled
reactors, on the other hand, can be compared with those achieved in the ad-
vanced converter reactors. The loss-of-coolant accident or the loss-of-flow
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vn-6
accident (more properly, depressurization) is a major consideration in the
safety analysis of reactors operating with pressurized coolant in that forced
circulation of coolant must be guaranteed. Nevertheless, it appears that
engineered safeguards could cope with realistic depressurization rates in the
gas- and steam-cooled reactors.
On the basis of the designs evaluated (Table III-6) and the combined
criteria of low power costs and good breeding capability, GCFR's have the
most potential of the concepts considered. Steam-cooled reactors suffer
either from higher power costs ( 1250 and 2680 psi SCBR's) or low breeding
ratio (3700 psi SCBR).
There is interest in the development of SCFR's because of the possibility
that breeding ratios comparable with those of liquid-metal-cooled fast breeder
reactors (LMFBR) can be obtained with a system utilizing the large background
of light water reactor technology and utility experience with steam systems.
The steam cooling technology is simpler than that for sodium. Since steam-
cooled reactors operate on a direct cycle, primary heat exchangers are not
required, and a simpler steam supply system results. Additional incentives
include the elimination of chemical reactions with the coolant, the possibility
of flooding the reactor with a transparent medium (water) during refueling
or maintenance operations, and the possibility of using water spray systems
for emergency cooling.
In the GCFR, helium gas at a pressure of from 70 to 100 atm is used to
transport the heat from the reactor core to the steam generators. Since the gas
does not become radioactive and cannot react chemically with the water in the
steam generator, there is no need for an intermediate heat exchanger. The
resulting simplification of the system is a helpful offset against the need to
design for a higher coolant pressure with gas.
The use of helium as a coolant has other special advantages for a fast
breeder reactor. Helium does not interact with the fast neutrons in the reactor
core, resulting in both simplified control of the reactor and enhanced breeding
of new fissionable fuel from fertile material. In addition, helium is transparent
and chemically inert, providing visibility during refueling and maintenance oper-
ations, a simpler engineering design and freedom from corrosion problems.
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VIII-1
VIII. EFFECT OF NUCLEAR POWER ON
NATIONAL AIR QUALITY
From the information used and developed in this report, it is clear that
an increasing preponderance of new steam generating stations will be nuclear-
fueled. During the next three decades, nuclear power plants will increase
from three percent to well over 40 percent of generating capacity, while fossil
units will increase on an absolute basis and decrease on a percent basis.
Fossil-powered plants supplied approximately 260, 000 megawatts of capacity
in 1970 and are forecasted to supply approximately 400,000 megawatts in 1980,
570,000 in 1990, and 740,000 in 2000. Typically, a large, relatively new
fossil plant operated at about a 64 percent load factor in 1970 (Ref. VIII-1).
This typical load factor can be expected to remain reasonably stable through
the next several decades although some fluctuations due to improved trans-
mission and dispatching system, greater plant reliability, and other techno-
logical and management developments may be expected. These factors, which
would generally tend to increase load factors, would be counterbalanced by
the increased use of nuclear plants for base-load and relegation of fossil
stations to load-following service.
From relatively scattered sources (Ref. VIII-2), nuclear plants may be
expected to operate generally at load factors in excess of 80 percent as base-
loaded facilities. With increasing numbers of nuclear plants coming into ser-
vice over the next 30 years, the nationwide average load factor may demon-
strate a gradual trend upward.
The effect of the use of nuclear power on fossil steam generating emis-
sions to the environment is of primary importance in the results of this study
effort. Therefore, extensive efforts were extended to derive reasonable fore-
casts of nuclear and fossil steam capacities over the period of interest. Data
from Figures V-2 and V-3 were used as a basis for the forecasts of total,
nuclear,and fossil steam capacities. The following points were considered:
• With minor exceptions, the forecasts of total electrical
power requirements through about 2020 were in quite good
agreement. Therefore, it was assumed that the midline of1
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VIII-2
Figure V-3 and the extrapolation of it represented total elec-
trical power capacity requirements. The values thus derived
are shown in Table VIII-1.
As shown in Figure V-2, there are substantial differences in
nuclear generating capacity forecasts, largely depending on
the year the projection was made. To provide some basis
for comparative purposes, it was assumed that the extreme
upper and lower values were too far off the midline curve to
be truly meaningful and representative and could be disregarded.
High, middle, and low values of nuclear capacity and the
extrapolations of these curves were then derived, excluding
extreme values. The results of this derivation are also
presented below in Table VIII-1.
It was assumed that fossil fuel usage never falls below 500,000
megawatts of capacity after 2000, since these plants may be
expected to be used continuously as peaking units and in iso-
lated areas with too low power demand for economical nuclear
generation.
From forecasted total capacity, and the high, middle, and low
values of nuclear power, the corresponding low, middle, and
high fossil capacities were derived. These values are included
in Table VII-1.
Based on proposed national performance standards for sta-
tionary sources promulgated by EPA (Ref. VIII-3), it was
assumed that the average fossil-fueled station will emit
approximately 10 pounds of sulfur dioxide per megawatt-hour,
corresponding to a one percent sulfur content of fuel (i.e., coal
and oil) and a mix of approximately 80 percent coal, 10 percent
oil, and 10 percent gas as fossil fuel throughout the time period.
The above emission value is approximately equivalent to 0. 8
pounds of sulfur dioxide per million Btu of liquid fuel and
1. 2 pounds of sulfur dioxide per million Btu of solid fuel.
This assumption does not allow for the fact that methods of
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TABLE VIII-1. CAPACITY FORECASTS BY FUEL,
1970
1980
1990
2000
2010
2020
2030
2040
2050
Total forecast steam
capacity (xlO^ Mwe)
270
500
930
1410
2310
3300
4500
5900
6900
Nuclear capacity
forecasts (x 10^ Mwe)
High (Case I)
10
190
650
910
1810
2800
4000
5400
6400
Middle (Case II)
10
105
350
700
1020
2000
3000
4200
5000
Low (Case III
10
50
120
240
470
650
900
1100
1400
Fossil capacity „
forecasts (xlO Mwe)
Low (Case I)
260
310
555
500
500
500
500
500
500
Middle (Case II)
260
395
580
710
1290
1300
1500
1700
1900
High (Case III)
260
450
810
1170
1890
2650
3600
4800
5500
TABLE VIII-2.
TONS OF S02
EMITTED PER YEAR (xlO6)
Case I (low fossil use)
8. 8
10. 8
14. 3
14. 3
14. 3
CO
14. 3
14. 3
Case II (middle fossil use)
11. 3
16. 5
20. 2
CO
CD
37. 1
42. 8
in
•
00
54. 2
Case III (high fossil use)
12. 8
23. 1
33. 4
53. 9
75. 5
102. 6
136. 8
156. 8
<
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VIII-4
reducing sulfur content of fuels (e.g., coal gasification) may
substantially reduce sulfur dioxide emissions. Using such a method,
emissions would be reduced proportionally to the amount of
sulfur removed. The important point is in the relative dif-
ferences caused by varying amounts of installed nuclear
capacity.
• A constant plant load factor of 65 percent was assumed
throughout.
Based on the primary data and the above-listed assumptions and premises,
three cases of fuel distribution were formulated. Case I assumes maximum
nuclear capacity and minimum fossil capacity. Case II assumes middle values
of both nuclear and fossil. Case III assumes minimum nuclear SLnd maximum
fossil capacities. In the table, the sum of Case I values equals the total fore-
casted capacity requirement. Similarly, the sums of Case II and Case III
values also equal total capacities. Finally, values of sulfur dioxide emissions
were derived, based on the assumptions outlined above. This information for
the three cases is shown in Table VIII-2 and, more graphically, in Figure VIII-1.
It should be again noted that the results shown in Table VIII-2 presume
no change in allowable sulfur dioxide emission rate through the period of
interest. Throughout the 80 years considered, sulfur dioxide emissions were
assumed to remain constant at 10 pounds per megawatt-hour. Any technological
or legislative changes which could affect fossil fuel sulfur dioxide emissions
would simply change the total emissions by a proportional amount. For example,
halving the allowable equivalent fuel sulfur content from one percent to one-half
percent would correspondingly reduce total emissions in Table VIII-2 by one-half.
From the results of the above analysis, it is significant to note the effects
of differences in nuclear power derived from the forecasts noted. For example,
in as short a time as 10 years, in 1980, the national emissions from fossil
steam generation could be slightly less than nine million tons per year if the
more optimistic forecasts of nuclear power supply are realized. On the other
hand, if the pessimistic (low) expectations of nuclear power supply are met,
the national sulfur dioxide emissions could be as high as 13 million tons per
year. This trend continues through the forecast period until the emission
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VIII-5
Figure VIII-1. Potential Sulfur Dioxide Emissions
From Fossil Steam Electric Plants
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VIII-6
difference in 2050 has reached over a factor of 10. Thus, the impact of the
successful development and use of the various forms of nuclear energy for
electrical power generation over the next 80 years plays a significant part
in air pollution control planning. At different development and use rates
over that time period, very large differences in sulfur dioxide emissions to
the atmosphere could occur.
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VIII-7
REFERENCES
VIII-1. 49th Semi-Annual Electric Power Survey, Edison Electric Institute,
New York, April 1971.
VIII-2. Derived from Chapter V references.
VIII-3. "Standards of Performance for New Stationary Sources," 42 CFR
466, Federal Register 36, 109, August 17, 1971.
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