PNL-3116.txt

PNL-3116
uc-20d

3 3679 00053 6153

FUSION-FISSION ENERGY SYSTEMS EVALUATION

Pacific Northwest Laboratory

V. L. Teofilo J. E. Morrison(b)
D. T. Aase R. T. Perry\C
W. E. Bickford A. D. Rockwood
B. R. Leonard gr. S. C. Schulte(d)
R. T. McGrathla C. E. Willingham

University of Washington

F. L. Ribe G. L. Woodruff
N. J. McCormick

January 1980

Prepared for
the U.S. Department of Energy
under Contract EY-76-C-06-1830

Pacific Northwest Laboratory
Richland, Washington 99352

(a) University of Michigan, Ann Arbor, MI

(b) Stanford University, Palo Alto, CA

(c) Now affiliated with University of Wisconsin,
) Madison, WI

(d) Now affiliated with CH2M Hill, Bellevue, WA
-
PREFACE

This report serves as the basis for comparing the fusion-fission (hybrid)
energy system concept with other advanced technology fissile fuel breeding
concepts evaluated in the Nonproliferation Alternative Systems Assessment
Program (NASAP). As such, much of the information and data provided herein
is in a form that meets the NASAP data requirements. Since the hybrid con-
cept has not been studied as extensively as many of the other fission
concepts being examined in NASAP, the provided data and information are
sparse relative to these more developed concepts. Nevertheless, this report
is intended to provide a perspective on hybrids and to summarize the findings
of the rather 1imited analyses made to date on this concept. This report was
developed jointly by Pacific Northwest Laboratory and the University of
Washington.
CONTENTS

PREFACE
FIGURES
TABLES
I. SUMMARY .
A. FUSION DRIVERS
B. FISSION BLANKETS
1. Once-Through Fuel Cycle
2. Pu Recycle
3. Refresh Cycle
4, U-233 Recycling
C. SECTION I REFERENCES
IT. INTRODUCTION . .
SECTION II REFERENCES
IIT. FUSION DRIVERS

Iv.

A.

E.

TOKAMAK
1. Plasma Physics .

2. Conceptual Engineering Design .

MIRROR
1. Plasma Physics .

2. Conceptual Engineering Design .

LINEAR THETA PINCH .
1. Plasma Physics .

2. Conceptual Engineering Design .

LASER INERTIAL
1. Inertial Fusion Physics

2. Conceptual Engineering Design .

SECTION III REFERENCES

FISSION BLANKETS

A.

B.
C.
D

FUEL FORMS

TRITIUM BREEDING MATERIAL CANDIDATES

COOLANTS . . .
HEAT TRANSFER - FLUID FLOW

vii
VI.

VII.

E
F
G.
H

STRUCTURAL DESIGN . .
MECHANICAL AND THERMAL HYDRAULIC DATA
REMOTE DISASSEMBLY AND MAINTENANCE
SECTION IV REFERENCES

NEUTRONICS

G M M O O W >

COMPUTATIONAL METHODOLOGY

NUCLEAR DATA

FISSILE FUEL BREEDING

TRITIUM BREEDING

BURNUP AND ISOTOPICS .
FISSILE FUEL AND POWER PRODUCTION .
SECTION V REFERENCES

CONCEPTUAL PLANT DESIGN

A.

B.
C.

PLANT LAYOUT .

1. Tokamak Hybrid Reactor

2. Mirror Hybrid Reactor

3. Laser Hybrid Reactor .
4, Linear Theta-Pinch Hybrid Reactor
POWER ANAYSIS .

SECTION VI REFERENCES

HYBRID FUEL CYCLE ANALYSIS

A.

FUELING ALTERNATIVES
1. No-Reprocessing

2. Reprocessing and Recycle of Fissile Materials

FUEL MANAGEMENT STRATEGIES
1. Tokamak Hybrid Reactor
2. Mirror Hybrid Reactor

FACILITY REQUIREMENTS

1. Fuel Fabrication - Mainline Process Description

2. U02/Pu02 Fuel Fabrication

3. U0
De

Refresh Fuel Cycle Fabrication
ThC Fuel Fabrication

cription

viii

§/PUO% Fuel Fabrication Mainline Process
VIII.

IX.

. Hybrid Fuel Storage .
Operational Waste Facilities
Reprocessing - Spent Hybrid Fuel

O 0 ~N O

Pu-Recycle to Thermal Reactor Reprocessing: Mainline
Process Descriptions . . .

10. Description of Process Steps

11. Thorex Process for U/Th Reprocessing in the Pu-Catalyst
Fuel Cycle . . . . . . .

12. Reprocessing Options
D. SECTION VII REFERENCES
PROLIFERATION RESISTANCE CONSIDERATIONS
A. INTRODUCTION - GENERAL CONSIDERATIONS
1. The Issue of Reprocessing
2. Fusion-Fission Reactors Studied

3. Fuel Cycle Operations of Interest for Non-
Proliferation

4. Standard of Comparison

NO REPROCESSING

REPROCESSING AND RECYCLING

1. Plutonium Recover and Recycle .

2. Denatured 233

U Cycle
3. High Gain Mixed Cycle
D. PROLIFERATION RESISTANCE ENGINEERING
1. Allowable Activities |
2. Proliferation Resistance Effectiveness Evaluation .
E. REFERENCES FOR SECTION VIII
ECONOMICS . . .
GROUND RULES AND ASSUMPTIONS .
CAPITAL INVESTMENT COSTS
BLANKET COSTS . .
ANNUAL OPERATING AND MAINTENANCE COSTS
FUEL CYCLE COSTS
LEVELIZED ENERGY COSTS
FISSILE FUEL VALUE .
MARKET PENETRATION .

— O Mmoo W >

ix

VII-57
VII-57
VII-57

VII-58
VII-58

VII-63

VII-63

VII-70
VIII-I
VITI-1
VIII-2
VIII-3

VITI-3
VIII-4
VIII-4
VIII-5
VIII-5
VIII-6
VIII-6
VIII-7
VIII-7
VITI-10
VIII-11
IX-1
IX-1
IX-1
IX-5
IX-5
IX-6
IX-7
IX-8
IX-10
J.

NONPROLIFERATION IMPACT .
NucTear Center .
“"Throw Away" Fuel Cycle

1
2
3. Co-processing
4. Refresh Blanket
5

. Denaturing
SECTION IX REFERENCES

X. LICENSING AND SAFETY

A‘
B.

GENERIC DISCUSSION OF THE HYBRID CONCEPT
GENERIC SAFETY AND LICENSING ISSUES

1. Radiation Exposure

2. Accidents .

TOKAMAK HYBRID. . . . .
1. Description of the Tokamak Hybrid Concept
2. Safety and Licensing Issues for the THR .
3. Liquid Lithium Spills

4. Magnet Safety

5. Criticality

6. Magnetic Fields

7. Cryogenics

8. Activation Products .

MIRROR HYBRID . . . . . .
1. Description of the Mirror Hybrid Concept.

2, Safety and Licensing Issues for the Mirror Hybrid

Reactor
THETA PINCH

1. Description of the Theta Pinch Hybrid Reactor

Concept

2. Safety and Licensing Issues for the Linear Theta

Pinch Hybrid Reactor
LASER FUSION HYBRID

1. Description of the Laser Hybrid Reactor Concept
2. Safety and Licensing Issues for the Laser Hybrid

Reactor

IX-12
IX-12
IX-15
IX-15
IX-16
IX-16
IX-17
X-1
X-1
X-3
X-3
X-6
X-11
X-11
X-12
X-14
X-15
X-15
X-16
X-17
X-18
X-19
X-19

X-20
X-24

X-24

X-24
X-27
X-27

X-28
XI.

X1I.

XIII.

G. SECTION X REFERENCES
ENVIRONMENTAL CONSIDERATIONS
A. FUSION FUEL CYCLE

1. Deuterium and Lithium
2. Tritium

3. Activation Products .
FISSION FUEL CYCLE .
MAGNETIC FIELDS

TOXIC LASER GASES

UNIQUE RESOURCE REQUIREMENT
SECTION XI REFERENCES

UTILITY AND INDUSTRIAL PERSPECTIVES - COMMERCIALIZING HYBRID
HYBRID REACTORS . .

A. SIGNIFICANCE OF COMMERCIALIZATION ISSUES
B. CONCEPTUAL MODEL OF THE COMMERCIALIZATION PROCESS
C. CHARACTERISTICS OF DEMAND

D. PROBLEMS OF PROPERTY RIGHTS

E. CAPACITY TO PRODUCE
F

G

H.

M m O O o

PRODUCT CHARACTERISTICS .
CONCLUSIONS
SECTION XII REFERENCES
TECHNOLOGY STATUS AND RD&D REQUIREMENTS
A. PRESENT STATUS OF FUSION PHYSICS
1. Tokamak
2. Mirror .
3. Linear Theta Pinch
4, Inertial Confinement
B. FUSION DRIVER RD&D REQUIREMENTS
1. Tokamak
2. Mirror
3. Linear Theta-Pinch
4, Inertial Confinement
5. ICF R&D Facilities

X1

X-32
XI-1
XI-1
XI-1
XI-2
XI-3
X1-4
X1-8
X1-8
X1-9
XI-11

XII-1
XII-1
XII-3
XI1I-6
X11-8
XII-12
XII-15
XII-17
XII-19
XTII-1
XTII-1
XIIT-1
XIII-4
XIII-7
XII1-7
XIII-13
XIIT-13
XITI-19
XIII-21
XITI-21
X1I11-27
C. PRESENT STATUS OF BLANKET ENGINEERING . . . . XIII-30

1. Neutronics Design . . . . . . . XIII-30
2. Thermal and Mechanical Design . . . . . XIII-31
D. BLANKET RD&D REQUIREMENTS . . . . . . XITI-35
1. Fission . . . . . . . . . XIII-35
2. Neutronics . . . . . . . XII1I-36
E. POSSIBLE HYBRID RD&D PROGRAM . . . . . . XIII-38
1. Program . . . . . . . . . XIIT1-38
2. Facilities . . . . . . . . XII1I-42
3. Funding Requirements . . . . . . XIII-43
F. SELECTION XIII REFERENCES . . . . . XI1I-45
APPENDIX A - CAPITAL INVESTMENT COST ESTIMATES . . . A-1
APPENDIX B - LEVELIZED ENERGY COST ESTIMATES . . . B-1

xii
[1-1
IIT-A-1
IIT-B-1

ITI-B-2
I11-B-3
III-B-4
I1I1-C-1

IIT-C-2

ITI-D-1
III-D-2

IV-C-1

IV-E-1
IV-E-2
IV-E-3
IV-E-4
IV-£E-5
IV-E-6
IV-E-7
IV-E-8
IV-G-1
IV-G-2
IV-G-3

IV-G-4

IV-G-5
V-A-1
V-C-2

FIGURES

Fusion-Fision Process
Cross-Section of the Tokamak Hybrid Reactor .

[1lustrating the Principles of a Magnetic-Mirror Device
in Minimum-B Geometry . . . . . . .

Overall View of the LLL-GA Mirror Hybrid Reactor .
Cutaway View of the LLL-GA Mirror Hybrid Reactor .
Mirror Hybrid Blanket Module .

ITlustrating the Principle of a Staged Theta-Pinch
Using Separate Shock-Heating and Adiabatic Compression
Coils . . . . . . .

Section of the Core of a Linear Fusion Reactor with the
Blanket Inside the Multiturn Compression Coils and
Shock Heating Coils . . .

The LLL-Bechtel 4000-MWt Laser-Fusion Hybrid Reactor

First Wall Structure of the LLL-Bechtel Laser Fusion
Hybrid Reactor . . . . . .

Thermal Efficiency of Typical Thermo-Dynamic Cycles as a

Function of Peak Cycle Temperature

Tokamak Hybrid Blanket Segment

Helium Coolant Flow in the Tokamak Hybrid Reactor
Cross-section View of the Tokamak Hybrid Reactor .
Tokamak Hybrid Module Detail .

Mirror Hybrid Blanket Submodule

Mirror Hybrid Blanket Arrangement .

Laser Hybrid Blanket Segment Arrangement

Linear Theta-Pinch Hybrid Blanket Module

Tokamak Hybrid Reactor Cross Section

Mirror Hybrid Reactor Blanket Module

Mirror Hybrid Reactor Blanket and Structural
Components

Alternate Blanket Replacement Technique for Mirror
Hybrid . . . . .

Mirror Hybrid Module Handling Machine
Reactor Calculational Schematic
UO2 Blanket Schematic

X111

I1-2
ITI-4

I1I-7

ITI-11
IIT-12
ITI-13

III-14

ITI-16
ITI-18

III-19

1v-8

IV-13
IV-14
IV-15
IV-16
IV-17
IV-18
Iv-20
IV-21
Iv-27
Iv-28

IV-29

IV-30
IV-31
V-1
V-6
V-C-3
V-C-4

V-F-1

V-F-2
V-F-3

VI-A-1]
VI-A-2
VI-A-3
VI-A-4

VI-A-5a

VI-A-5b

VI-A-6
VI-A-7
VI-B-1
VI-B-2
VI-B-3
VI-B-4
VII-A-1
VII-A-2
VII-A-3
VII-A-4
VII-A-5
VII-A-6
VII-B-1

VII-B-2
VII-B-3

VII-B-4

VII-B-5

UC Blanket Schematic
Pu02-ThC2 Blanket Schematic

Heating Rates as a Function of Radius for Three Blanket

Types

A Fast and Thermal Group Flux as a Function of Radius .

A Fast and Thermal Group Flux as a Function
of Reactor Radius

Tokamak Hybrid Reactor Hall

Power Conversion System for Tokamak Hybrid Reactor
Mirror Hybrid Plant Layout

Mirror Hybrid Reactor

Mirror Hybrid Coolant Systems (Primary Coolant
Loop) . . . .

Mirror Hybrid Coolant Systems (Secondary Coolant
System . . . . . .

Laser Hybrid Reactor Building Layout
Linear Theta-Pinch Hybrid Reactor Configuration
Tokamak Hybrid Plant Schematic
Theta-Pinch Hybrid Plant Schematic
Laser Hybrid Plant Schematic .
Mirror Hybrid Plant Schematic
Uranium Nuclear Fuel Cycle
Once-Through Hybrid Fuel Cycle
Refresh Hybrid Fuel Cycle

Thorium LWR Fuel Cycle

Thorium Hybrid Fuel Cycle

Plutonium Recycle

Tokamak Hybrid Reactor Fuel Flow Once- Through Fuel
Cycle .

Tokamak Hybrid Reactor Fuel Flow - Pu-Recyc]e

Tokamak Hybrid Reactor Fuel Flow - Pu- Catalyst Fuel
Cycle . . . .

Tokamak Hybrid Reactor Fuel Flow - Refresh Fuel
Cycle . . . .

Mirror Hybrid Reactor - Pu-Catalyst

Xiv

V-7
V-8
VII-B-6
VII-B-7
VII-B-8
VII-B-9

VII-B-10
VII-C-1
VII-C-2

IX-F-1
XIT-A-1
XIT-B-1

XITI-A-1
XITI-A-2

XITI-B-1
XII1-B-2
XIIT-B-3
XIII-B-4
XITI-E-1

Laser Hybrid Reactor - Once-Through Fuel Cycle
Laser Hybrid Reactor - Pu-Recycle .
Laser Hybrid Reactor - Pu-Catalyst

Theta-Pinch Hybr1d Reactor - Pu- Recyc]e to Thermal
Reactor . . .

Theta-Pinch Hybr1d Reactor - Pu-Cata]yst
Fabrication Facility Layout . . .
/Pu0 Fabrication Fac111ty for Pu- Cata]yst Fuel

Cyg .

Annual Cost of E]ectr1c1ty and Levelized Energy Cost
Scope of Commercialization

Conceptual Model of Commercialization

Technical Progress and Outlook in Magnetic Fusion

Types of Laser Pellets as Projected by Lawrence Livermore
Laboratory . . .

Major Facilities Schedule

Engineering Facilities Schedule

Tandem Mirror Reactor

Inertial Confinement Fusion Facilities Schedule
Hybrid Development Facilities Schedule .

Xy

VII-37
VII-38
VII-39

VII-45
VII-46
VII-50

VII-53
IX-9
XII-3
XII-4
XIII-3

XIII-12
XIII-14
XIII-17
XIII-20
XITI-29
XII1-40
I-A-1
I-B-1
[-B-2
I-B-3
I1I-A-1
ITI-A-2
IV-B-1
IV-D-1
IV-F-1

IV-F-2

IV-F-3

IV-F-4

V-C-1
V-E-1
V-F-1
VI-B-1
VI-B-2
VI-B-3
VI-B-4
VII-A-1
VII-B-1
VII-B-2
VII-B-3
VII-B-4
VII-B-5
VII-B-6
VII-B-7
VII-B-8
VII-B-9

TABLES

Fusion Driver Characteristics . .
Once-Through/Plutonium Breeding Hybrids .
Fuel Refreshing Hybrids

233U Breeding Hybrids

Plasma Parameters for Tokamak Hybrid Reactor .
Power Requirements for a Tokamak Hybrid Reactor
Breeding Compound Characteristics

Typical Reactor Power Densities

Tokamak Hybrid Mechanical and Thermal Hydrau11c
Information .

Mirror Hybrid Mechanical and Thermal Hydrau11c
Information . . . . .

Linear Theta-Pinch Mechanical and Thermal Hydraulic
Information . . . .

Laser Hybrid Mechanical and Thermal Hydraulic
Information . . .

Blanket Neutronic Characteristics

Isotopic Concentrations After One Year Operation
Blanket Fissile Fuel and Fission Power Production .
Tokamak Hybrid Plant Parameters

Theta-Pinch Hybrid Plant Parameters

Laser Hybrid Plant Parameters .

Mirror Hybrid Plant Parameters.

Driver/Blanket Fuel Cycle Combinations

Tokamak Hybrid Fuel Management Data

Reactor Subsystems

Tokamak Hybrid Reactor Initial Material Requirements
Once-Through and Pu-Recycle Fuel Charge Data .
Once-Through and Pu-Recycle Fuel Discharge Data
Pu-Catalyst Fuel Charge Data

Mirror Hybrid Fuel Management Data .

Mirror Hybrid Reactor Initial Material Requirements
Laser Hybrid Fuel Management Data

Xvii
VII-B-10
VII-B-11
VII-B-12
VII-B-13
VII-B-14
VII-B-15
VII-B-16
VII-B-17
VII-B-18

VII-B-19
VII-B-20
VII-B-21
VII-C-1

VII-C-2
VII-C-3
VII-C-4

VIII-D-1
IX-A-1
IX-B-1
IX-B-2
IX-D-1
IX-E-1
IX-F-1
IX-G-1
IX-H-1
IX-H-2

IX-H-3
IX-H-4
X-C-1

Pu-Recycle Fuel Charge Data

Pu-Recycle Fuel Discharge Data

Pu-Catalyst Fuel Charge Data .

Laser Hybrid Reactor Initial Material Requirements
Once-Through and Pu-Recycle Fuel Charge Data
Once-Through and Pu-Recycle Fuel Discharge Data
Pu-Catalyst Fuel Charge Data . .
Theta-Pinch Hybrid Fuel Management Data

Theta-Pinch Hybrid Reactor Initial Material
Requirements . .

Once-Through and Pu-Recyc]e Fuel Charge Data
Once-Through and Pu-Recycle Fuel Discharge Data .
Pu-Catalyst Fuel Charge Data .

Once-Through and Pu-Recycle to Thermal Reactors Fuel
Fabrication Facility

Pu-Catalyst Fuel Fabrication Fac111ty
Characteristics

Reprocessing Facility Summary Data for Pu-Recycle
Fuel Cycle . . . .

Reprocessing facility Summary Data for Pu-Catalyst
Fuel Cycle . . .

Methods of Spiking Plutonium .

Economic Parameters/Unit Costs

Capital Investment Cost Summary

Capital Investment Cost Summary

Annual Operating and Maintenance Cost Summary
Fuel Cyclie Cost Summary

Levelized Energy Cost Summary

Fissile Fuel Breakeven Values

Energy Supply Scenarios .

Market Penetration Assessment Economic and Performance
Parameters

Market Penetration Assessment/Scenario 1
Market Penetration Assessment/Scenario 2
Engineered Safety Features for Fusion Magnets

xviii

VII-29
VII-30
VII-31
VII-33
VII-34
VII-35
VII-36
VII-40

VII-41]
VII-42
VII-43
VII-44

VII-51

VII-56

VII-65

VII-68
VIII-9
IX-2
IX-4
IX-4
IX-6
IX-7
IX-8

IX-10

IX-10

IX-11
IX-13
IX-14

X-16
XI-E-1
XI-E-2

XIII-A~1]
XITI-A-2
XIII-A-3
XITI-B-1
XIII-B-2

XIII-B-3
XIII-B-4
XIII-B-5
XIIT-E-1
XIII-E-2

Total Domestic Demand for Important Fusion Materials

Depletion of 1974 U.S. Reserves of Important Fusion
Power Plant Materials, Assuming No Additions to
Reserves . . . . .

U.S. Tokamak Research

U.S. Tokamak Experiments . . .

Office of Laser Fusion Physics Through Mid 1980s
Objectives of Major Fusion Reactor Facilities

Features of the Tokamak Fusion Driver Related to Large
Tokamak Experience . . . . .

Objectives of Major Fusion Engineering Facilities .
Sandia Accelerators . . .

Objectives of Major ICF Facilities .

Hybrid Reactor Technological Advance Requirements .
Hybrid Facility Cost Estimates

XX

XI-10

XI-10
XITI-1
XIIT-2
XIII-9
XIII-15

XITI-16
XIII-18
XIII-24
XIII-28
XIII-40
XIII-44
I. SUMMARY

The Office of Fuel Cycle Evaluation of the Department of Energy is con-
ducting a Nonproliferation Alternative Systems Assessment Program (NASAP).
The goal of the NASAP is to provide recommendations in the development of
nuclear energy systems which have potential for reducing the risk of nuclear
weapons proliferation while satisfying the short- and long-term needs for
nuclear energy. The fusion-fission hybrid is one of the nuclear enerqy
systems which have been considered for long-term applications. This report
represents the development of the information and data needed to evaluate
and analyze hybrids for the NASAP. Although most of the combined driver-
blanket hybrid systems considered in this study have not been optimized for
performance and cost, the resulting data provides valuable insights of the
future prospects and potential of hybrid development.

A.  FUSION DRIVERS

The fusion driver reactor systems with available information for both

inertial and magnetic confinement have been reviewed and analyzed. These
systems have been subjected to a preliminary screening whereby they have

been assessed in terms of electrical energy self-sufficiency; fuel production
to support a sufficient number of fission burner converters; acceptable neutron
wall loading and/or blanket power density; and scientific and technological
feasibilities. Some of the characteristics of those driver systems which have
been retained for evaluation in this study are listed in Table I-A-1.

These systems include the laser heated inertial confinement hybrid with

(1)

high gain pellets based upon the Lawrence Livermore Laboratory-Bechtel Study 5
the Tokamak operated in the ignition mode designed by PNL and based upon the

(2); the classical mirror with Yin-Yang

(3)

Tokamak Demonstration Hybrid Reactor
magnets based upon the LLL-General Atomic hybrid design‘*~’; and a linear theta-
pinch designed by the University of Washington. These systems generate fusion
power of 400-1100 MW with neutron wall loadings of 1-2 Mw/mz. When combined
with selected fission fueled blankets, they provide a neutron economy which
may prove advantageous in the production of proliferation resistant fuel forms

and/or in situ fissile fuel burning.

I-1
TABLE I-A-1. Fusion Driver Characteristics

LASER IGNITED CLASSICAL LINEAR
INERTIAL TOKAMAK MIRROR ©-PINCH

REACTOR CAVITY DIMENSIONS (m)  10x13.4 1.2x5.4 8.0 0.6x500
FUSION GAIN 250 30 0.67 6.5
nT(s/m?) 1020 1020 2.3x1019 1020
HEATING POWER (MW) 3.4 .10 630 170
FUSION POWER (MW} 850 1140 404 1090
NEUTRON WALL LOADING (MW./m2) 1.8 2.2 1.6 0.9

B. FISSION BLANKETS

Previous hybrid blanket designs have generally proven to be undesirable
from the nonproliferation viewpoint simply because most of them were guided
by the desire to produce plutonium or U-233 without consideration of nonpro-
liferation issues. With this information new blanket concepts having perceived
nonproliferation advantages have been combined with the above fusion driver
systems and the resulting hybrids and their associated fuel cycles have been
characterized. A generic modular designed blanket was selected consisting of

a stainless steel structure. It contains regions for stainless steel clad
fertile fuel in addition to L1'02 for tritium breeding. The fuels are cooled with
high pressure helium. An appropriate number of such modules have been

designed to fit in the blanket region of each driver system, Different

fertile fuels were used for the characterization of four fuel cycles.

1. Once-Through Fuel Cycle

Using natural uranium carbide as the fuel in the fertile region of the
blanket, the fuel cycle can be either a once-through "throwaway blanket"
cycle, in which the fissile fuel is burned in situ, or it can be used to
breed fissile plutonium fuel to be used in fission reactors. The throwaway
blanket concept is analogous to the LWR once-through system with verified
spent fuel storage. The hybrid would only produce electric power for sale
and its spent uranium fuel would be cooied and shipped to a secure repository
for storage and ultimate disposal. Compared with the LWR once-through fuel

I-2
cycle, the hybrid "throwaway" blanket eliminates the need for enrichment
requirements, but it still requires similar safeguardé for the spent fuel.
It has markedly improved resource utilization since natural or even depleted
uranium could be used. However, with the present fusion driver concepts it
appears to be economically inferior to LWRs since it involves plants with
significantly greater capital costs to the extent that it would at least
triple the cost of the electricity produced as noted in Table I-B-1.

2. Pu Recycle

If the plutonium fissile fuel of the same throw-away blanket is recycled
to LWRs, the combination of the above blanket with the fusion driver systems
yields hybrids having the performance characteristics as listed in Table I-B-1.

TABLE I-B-1. Once-Through/Plutonium Breeding Hybrids

Laser Ignited Classical Linear
Inertial  Tokamak Mirror 6-Pinch
Thermal Power (MWt) 3300 4150 2580 4835
Net Electric Power (MWe) 940 1000 140 45
Blanket Fuel ucC uc uC uc
Pu Production Rate (kg/yr) 1325 1950 810 2590
LWR Support Ratio 4.0 5.8 2.4 7.8
Recirculated Power Fraction 0.24 0.29 0.89 0.98
Capital Cost ($/kWt) 617 501 997 531
Incremental Energy Cost
(A System Cost/LWR Pu-Recycle)
Pu Br: 0.34 0.20 1.00 1,50
Once-Through: 2.4 2.1 25.7 144 .3

These are the same characteristics for the once-through cycle except for the
reduction of the incremental energy cost of producing electric power with the
hybrid integrated with the LWRs it supports. This incremental cost is the
percent increase over the cost of electric power produced by LWRs which recycle
their own plutonium supplemented by the plutonium produced by a fast breeder
reactor. It is aslittle as 20% to 34% for the high Q drivers (tokamaks and
lasers) and as high as 100% to 150% for the low Q drivers (mirrors and e-pinch).
For enhanced proliferation resistance of the recycled plutonium fuel cycle,

I-3
the reprocessing and fuel fabrication facilities could be located in an Inter-
national Nuclear Center (INC) where co-processing and/or spiking of the final
LWR fuel would be performed. The hybrid need not be located in the INC unless
it contained an initial inventory of fissile fuel. The resource utilization

of this cycle is favorable since use can be made of unenriched or even depleted
uranium as well as the recycled plutonium. The hybrid system is economically
attractive with this fuel cycle because of the large number of fission reactors
which it could support.

3. Refresh Cycle

The fuel refreshing hybrid cycle utilizes natural uranium oxide fuel in
the fertile regions of the blanket modules. 1In this refresh cycle, after the

UO2 fuel is enriched in the hybrid blanket to the necessary level, it is reused
in LWR systems after appropriate mechanical recladding and reassembly compatible
with LWR systems. After the fuel is burned and its fissile content depleted in
the LWR system, it may be again reclad and reassembled for refreshing or re-en-
riching in the hybrid. The performance characteristics of the resultant driver-
blanket combinations are listed in Table I-B-2. For this cycle, only the ignited
tokamak hybrid provides the necessary 14 MeV neutron fluence and initial inven-
tory to allow for a practicable time (4 years) for enrichment of the UO2 fuel

to the 3% level. In that case the system economics indicate an incremental

cost of electric energy slightly above the cost for plutonium recycling using
the same hybrid system. It has the advantage of resource utilization since
natural or depleted uranium or even thorium could be used and its nonprolifer-
ation attractiveness rests on the fact that no chemical reprocessing is involved
in this fuel cycle.

4. U-233 Recycling

Perhaps the potentially most attractive hybrid blanket concept is one in
which a zone of natural uranium oxide in an equilibrium mixture with recycled
plutonium oxide is used for neutron and energy multiplication to enhance the
production of U-233 in a natural thorium carbide fueled region. The recycled
U-233 can then be denatured with U-238 and used in fission reactors to enhance
proliferation resistance. This concept has high resource utilization since it
makes use of thorium and recycled U-233 which can produce relatively high con-
version ratios in thermal fission reactors. It also incorporates the superior

I-4
performance of U-238 and recycled Pu-239, As seen in Table I-B-3, the perform-
ance characteristics of the hybrids fueled with such a blanket concept indicate
its economics may be superior to any of the other fuel cycles since it could

produce the most power and fuel when combined with the same driver systems.

TABLE 1-B-2, Fuel Refreshing Hybrids
LASER IGNITED CLASSICAL LINEAR
INERTIAL TOKAMAK MIRROR B8-PINCH
THERMAL POWER (MWht) 3015 37156 2400 4350
NET ELECTRIC POWER (MWe) 830 853 70 -175
BLANKET FUEL uo, uo, uo, uo,
Pu PRODUCTION RATE (kg/yr) 940 1390 575 1845
LWR SUPPORT RATIO 2.8 4.2 1.7 5.5
RECIRCULATED POWER FRACTION 0.27 0.32 0;94 ;].09
CAPITAL COST ($/kWt) 544 536 1038 546
INCREMENTAL ENERGY COST — 0.22 — —
(ASYSTEM COST/LWR Pu-RECYCLE)
TABLE I-B-3. 233y Breeding Hybrids
LASER IGNITED CLASSICAL LINEAR
INERTIAL TOKAMAK MIRROR 6-PINCH
THERMAL POWER (MW1t) 4980 6600 3600 8200
NET ELECTRIC POWER (MWe) 18670 1835 545 1560
BLANKET FUEL ThC ThC ThC ThC
(Pu0,-U0,;} (Pu0O,-UO,} (Pu0,-U0,) (PuO,-U0,)
233 PRODUCTION RATE (kg/yr) 2585 3810 1575 5070
LWR SUPPORT RATIO 9.5 14 5.8 18.6
RECIRCULATED POWER FRACTION 0.16 0.17 0.67 0.58
CAPITAL COST ($/kWt) 557 396 830 463
INCREMENTAL ENERGY COST 0.14 0.04 0.42 0.26

(ASYSTEM COSTS/LWR Pu-RECYCLE)

I-5
In addition to systems design, resource utilization and economics, the
hybrid systems which have been analyzed and characterized in this study have
also been evaluated on a normalized basis with respect to safety and environ-
mental factors, proliferation resistance, commercialization, as well as tech-
nological requirements. With the very significant absence of criticality as
a key concern, the hybrid introduces no issues which have not been identified
in the fission and fusion programs. Because it is the earliest proposed com-
mercial application of fusion energy, the hybrid may be the first energy
systems to introduce the unique fusion issues (e.g., tritium management, vacuum
rupture, magnet accidents) to the licensing community. This is not seen as
time-constraining on the date for introducing the first commercial systems
providing the identified issues are resolved without delay.

An analysis has been done on the nonproliferation aspects of the hybrid
and its associated fuel cycles relative to fission reactors. It is evident
that any fission fuel cycle option recommended for reduced pro]iferation can
be adopted with hybrids in the system. Moreover, new fuel cycles can be

envisioned which start with natural or depleted material and discard the spent
fuel elements. However, these may be unacceptable from an economic standpoint.

The utility and industrial perspectives on hybrid reactors are examined
within the context of the commercialization process. Specific issues in the
process are identified and reviewed for the case of hybrid reactor concepts.
This illuminates the key factors which will influence private sector's deci-
sions to invest in fusion-fission reactors. 1In addition, some of the public
decision-making problems are highlighted.

The required level of technology for both the fusion and fission components
of a commercial hybrid system are technologically feasible. The fusion-side
scientific and technological performance requirements are perceived as being
attainable as a next step following the current generation of confinement experi-
ments (c. 1985). Similarly, the fission-side requirements are perceived as
having been demonstrated or could be demonstrated with a modest investment of
research and development funds. A possible hybrid facilities development sched-
ule has been developed which allows for the parallel development of both mag-
netic and inertial fusion drivers as well as hybrid blankets. Such a schedule

I-6
would allow the driver selection to be made by 2000 for the first economically
prototypical hybrid reactor which conceivably could operate as early as 2010.

1-7
C.

SECTION I REFERENCES

Bechtel Corp., Laser Fusion-Fission Reactor Systems Study. UCRL-13796,
University of California Research Labs, Livermore, CA, July 1977.

R. W. Conn, et al., "TDHR - A Tokamak Demonstration Hybrid Reactor,"
TANSA, 27(26), 1977.

D. J. Bender, et al., Reference Design for the Standard Mirror Hybrid
Reactor, UCRL-52478, Lawrence Livermore Laboratory, Livermore, CA, May
1978.

I-8
IT. INTRODUCTION

A concept which has potential for future application in the electric power
sector of the U.S. energy economy is a combination of fusion and fission tech-
no]ogy.(1) The fusion-fission energy system, called a hybrid, is distinguished
from its pure fusion counterpart by incorporation of fertile materials (uranium
or thorium) in the blanket region of a fusion reactor.

The neutrons produced by the fusion process can be used to produce fuel
for fission power reactors through capture events in the fertile material. For
the current hybrid design concepts being studied, it is expected that 5 to 15
Light Water Reactors (LWRs) of 1000 MWe capacity can be supported from the
annual fissile production from the hybrid. Although fuel production is envi-
sioned as the chief benefit of a fission-fusion system, the thermal energy
generated through fission events in the blanket could be used to generate
electricity. The fact that hybrid reactors could produce power as weli as
fuel to extend the fuel supply for fission reactors has been the subject of
many studies(z). ”Thése studies have shown that fuel-producing hybrids capa-
ble of fueling multiple burner-converters can serve a useful function in the
perceived market place shortly after the year 2000. However, they conclude
that hybrid breeders must produce and sell power at least sufficient to offset
the power consumed by the devices in order to compete in the marketplace. The
sale of fissile material probably requires chemical processing of the blanket
to recover the fuel, although recycle without reprocessing has been suggested

(3)

The hybrid may be able to play multiple roles in the nuclear power economy.
Projections of the electric generation mix in the U.S.,(4) to the year 2000,
predict a potential shortfall of fissile material shortly after the year 2000.
Interest in hybrids therefore stems from the possibility that fuel breeding
hybrids might be developed and deployed in time to ease or eliminate this
potential shortfall and stabilize fissile fuel costs. In addition, because of

235U, electrical utilities relying on

the uncertainty in the future supply of
nuclear power are interested in the hybrid concept to produce fissile fuel for
existing power plants. With an additional supply of fissile material, the

future nuclear increment of the electric generation mix might grow substantially.

In the fusion-fission reactor, as depicted in Figure II-1, the 14 MeV fusion
neutron deposits its energy in the blanket where it is absorbed by the fertile

II-1
- WORKING FLUID

BLANKET

PLASMA

FISSION
FRAGMENT

FISSION FRAGMENT

ENERGY RELEASED 200 MeV

TO ENERGY -
CONVERSION SYSTEM

FIGURE II-1. Fusion-Fission Process

material. Subsequent reactions. neutron reemission, fission or capture, can
take place depending upon the energy of the absorbed neutron. If the incident
neutron energy is greater than 12 MeV, the neutron multiplying reactions

(n, 2n) and (n, 3n) as well as the fission reactions with 238 and °3°Th are
dominant. If the Neutron energy is degraded below ~2 MeV, the principal
absorption reaction is radiative capture (n, y) in the fertile fuel.

Through subsequent decay, the end products are the isotopes 239Pu or 233

u.
These isotopes are both fissile materials and thereby candidate fuels for
fission power plants. In addition, neutron reactions with the isotopes of
1ithium in the blanket will absorb or yield energy, depending upon the iso-
topic content, and produce tritium for replenishing the T supply consumed in
the fusion process.

In comparing the fusion process with the process in a hybrid, it should
be noted that more energy is released in the hybrid. The fusion process
yields 18 MeV of energy whereas fission in the hybrid blanket yields ~180 MeV,
roughly ten times more energy release. In the high energy absorption and
fission processes, additional neutrons are also released. Thus, in the hybrid
both energy as well as neutron multiplication take place. This may be consid-
ered a desirable feature for reactor applications. The power output require-
ments of the fusion driver may be reduced compared to the pure fusion system
for producing the equivalent amount of electric power. Thus, the performance
requirements of the fusion driver component may be somewhat less stringent

than those for pure fusion electrical power plants. This difference is prob-
ably small for fusion driver concepts with attainably high fusion gains (Q>20).
However, for those fusion confinement concepts with achievably low gain (Q<20),
conceptual studies have indicated that the fusion component performance require-
ments are substantially lower for the hybrid than for its pure fusion counter-
part.

| The major fission technology requirements for the hybrid are expected to
be developed in the course of research and development of fission power
reactors and their fuel cycles. Those fission components needing development
require only a modest incremental investment of research and development funds.
In addition, the fission blanket is inherently subcritical which precludes
criticality accidents and mitigates the afterheat problems suffered in potential
loss of coolant accidents compared to similar events in LWRs.

The hybrid concept may be a viable supplement or alternative to the LMFBR
to extend the nuclear energy option beyond the next century. It may also be
looked upon as a step along the pathway to pure fusion power. It is conceiv-
able that many uncertainties in plasma physics, plasma engineering, and blanket
engineering performance of pure fusion systems could be resolved through the
development of hybrids. Thus, hybrids would be a step on the road to achieving
the benefits of pure fusion technology. With the present schedule of develop-
ment of fusion as well as fission technology, it is conceivable that a hybrid
could be developed near the turn of the century.

In this report the selected fusion driver concepts with proposed blanket
designs and their associated fuel cycles have been characterized. In addition
to a detailed economic analysis of these hybrids, related issues on prolifer-
ation resistance, safety and environment and commercialization are presented.
The technology status and RD&D requirements of the related technologies are
reviewed and a proposed hybrid RD&D program is presented.

II-3
SECTION IT REFERENCES

1. B. R. Leonard, Jr., "A Review of Fusion-Fission (Hybrid) Concepts."
Nucl. Tech. 20:161, December 1973.

2. R. C. Liikala, R. T. Perry, V. L. Teofilo, and B. R. Leonard, Jr.,
Perspectives of the Fusion Fission Energy Concept. PNL-SA-6492, Pacific
Northwest Laboratory, Richland, WA, February 1978.

3. R. L. Engel, and D. E. Deonigi, Evaluation of Fission-Fusion Hybrid
Concepts, Part A. ER-469, Project 268-1, Electric Power Research
Institute, Palo Alto, CA, 1976.

4. Research and Development Program for 1978-1982 Overview, Electric Power
and Research Institute, Palo Alto, CA, September 1, 1977.

11-4
IIT. FUSION DRIVERS

The fusion driver reactor systems with available information in the
literature for both inertial and magnetic confinement have been reviewed
and analyzed. These systems have been subjected to a preliminary screening
whereby they have been assessed in terms of electrical energy self-sufficiency;
fuel production to support a sufficient number of fission burner converters;
acceptable neutron wall loading and/or blanket power density; and scientific
and technological feasibilities. Those systems which have been retained
in this study and are described in this section include an ignited tokamak,
a classical mirror, a linear theta pinch with end plugging, and laser
inertial confinement system with high gain pellets.

A.  TOKAMAK

1. Plasma Physics

The fusion core of a Tokamak Hybrid Reactor (THR)(]) should have the
highest possible fusion power density to maximize the neutron fluence sup-
plied to the surrounding fusion blanket. In a Two-Energy Component Tokamak
(rct), (2)
transport and radiation losses by means of injected energetic deuterons

the temperature of the tritium bulk plasma is maintained against

which undergo fusion reaction with the relatively cold tritons. At plasma
temperatures <10 keV the maximum fusion power obtainable this mode of
operation is considerably larger than that obtainable for an ignited plasma
composed of a 50/50 D-T mixture. However, operation in the TCT mode

requires that the neutral beam injectors remain at full power during the
entire burn. This places strict performance requirements on the neutral

beam system and, more importantly, demands that a sizable recirculating

power fraction be maintained to meet the large power requirements for
continuous operation of the beam injector system. Considering these factors,
the desired fusion power level for the THR is obtained by using a >10 keV
ignited 50/50 D-T plasma. This relaxes the performance demands on the

neutral beam system and establishes an efficient operating cycle by minimizing
the recirculating power requirements. Under ignition conditions the plasma
temperature is maintained by the confinement of fusion alpha particles which
is sufficient to balance the transport and radiation losses.

ITI-1
High toroidal field, high beta and elongated plasma cross sections
are found to be essential for obtaining an ignited tokamak plasma. The
ignited plasma which was designed for the tokamak driver has the
characteristic parameters as listed in Table III-A-1.

A cross-sectional view of the Tokamak Hybrid Reactor is shown in
Figure III-A-1. The plasma cross-section is in the shape of a flattened
"D" (S = 1.53). This cross-section lends itself well to the implementa~
tion of a double-null poloidal divertor, which is used for the removal
of D and T ions, impurities, and alphas emerging from the discharge. The
elongated plasma cross-section, however, has a negative decay index; hence,
feedback stabilization of the plasma vertical position is required.

2. Conceptual Engineering Design

The first or vacuum wall of the THR consists of a 0.5 cm carbon liner
inside a double-walled stainless steel shell 5 cm thick having channels
for helium coolant at 700 psi. On the inner zone of the torus, where a large
fraction of the tritium is bred, the stainless steel backing is 1.5 cm
thick. This carbon-stainless steel first wall will be subjected to a
neutron flux of 2.2 Mw/mz. The radiation to the first wall is approximated
to be on the order of 25 MW resulting in a heating rate of 5.9 W/cm2. This,
together with the neutron flux, will result in a heating rate in the first
wall region of approximately 60 w/cmz. The coolant flow rate through the
first wall coolant channels of 190 kg/s at 70 m/s velocity provides a heat
transfer coefficient of 0.35 W/cm2-°C which is sufficient to keep the
first wall at 35°C.

This can be provided by 110 rotatable cryo-sorption pump pairs, 55
in each divertor zone, similar to those designed for the Tokamak Engineering

(3)

time. As soon as the cryo-sorption surfaces of the on-1ine pump are saturated,

Test Reactor. One-half of these pumps are to be on-line at any given
the pump pair is rotated 180°, placing the freshly regenerated pump in
place to begin pumping.

The toroidal field magnet system consists of 20 cryogenically stable
superconducting coils with Nb3Sn filaments in OFHC copper stabilizer.
The TF coils are constant tension "D" shaped, which produces a magnetic field
of 6.66 T on the plasma axis. The formulation corrects the magnetic forces

I1I-2
TABLE IIT-A-1. Plasma Parameters for Tokamak Hybrid Reactor

RO 5.4 m

a 1m

A 5.4

Elongation, K 2.0

Shape Factor, S 1.53 (flattened "D" shape)
Horizontal Wall Radius 1.3 m

Wall Area 424 m2

Plasma Volume 175 m3

Axial Bt 6.66 T

Ip 5.6 MA

q 2.4

fiy 2.54 x 101 4em™3
T,=T, 11.5 keV

fieTE 4.2 x 10 %em3s
Bp 3.8

Fusion Power 1160 MW

Neutron Power 928 MW

Neutron Wall Loading 2.2 MW/m2

Power Density 6.6 Mw/m3

ITI-3
BEAM PORT

SCCCCRN

m/

N

///

Section of the Tokamak Hybrid Reactor

Cross

FIGURE III-A-1.

I11-4
for variations in field due to the discreteness of the finite number of
coils and the shape of the cross section resulting in field ripple at the
plasma surface of only 1%. Conductors embedded in structural discs are
employed in order to hold the conductor rigidly within the supporting
structure.

Charged particles leaving the plasma are guided along the magnetic
field lines into the poloidal divertor zone where they give up their
energy by striking a sacrificial plate and are then pumped out in molecular
form. The divertor entrance width is set at 30 cm to keep the plasma
capture efficiency close to unity. Backflow of neutrals into the torus
must be prevented in order to minimize charge-exchange loss of fast ions,
as well as charge-exchange neutral sputtering. The required cryogenic
pumping surfaces can be readily accommodated inside the large TF coils.
The plasma flow in the scrapeoff region proceeds nearly at the speed
of sound the density here is relatively low (~2 X 10]3 cm“3), and the
plasma temperature is high (~2 keV).

Neutral beam injectors will be used to heat the THR plasma during
startup. Positive and negative ion source systems were considered for
the neutral beam injectors. For the THR beams at 150 keV, tolerable
net electrical efficiency can be obtained easily with positive ions,
provided direct conversion is employed to recover most of the power in
the unneutralized beam fraction.

The injector system for THR neutral beam heating is a 1980's technology
positive ion system. Twelve beam lines, each containing seven positive ion
sources arranged in a vertical array, will be used to deliver 150 MW to the
plasma. At the first wall each beam line fills a window 96 cm (horizontal)
by 25 c¢cm (vertical). The beam ports take up less than 1% of the first wall
area. To provide 150 MW of power to the plasma at 150 keV, an injection
current of 1000A equivalent is required.

Table III-A-2 lists the power requirements for the THR. As seen, a
recirculating power of 410 is needed. This corresponds to a plant
efficiency of about 70%.

ITI-5
TABLE III-A-2. Power Requirements for a
Tokamak Hybrid Reactor

MW
Helium Circulating Pumps 175
Cryo-generation Systems 70
Resistive Loss of VF Coils 100
Resistive Loss of OH Coils 140
Divertor Requirements 6
Feedback Stabilization System 30
Cooling Towers 10
Neutral Beam Requirements 5.15
Additional System Support 13.85
Total 410

B. MIRROR

1. Plasma Physics

Because it is an open-end device with an intrinsic loss of plasma,
the magnetic mirror does not admit operation at high Q values approaching
those of ignition. Under ideal circumstances the theoretical value of Q
for the plasma is only slightly greater than unity. The magnetic-mirror
reactor is therefore a driven power amplifier whose thermonuclear power
output is a factor of Q times its injected power. In order to achieve
economical net electrical output with such low values of Q, a magnetic-mirror
reactor must use the plasma energy which escapes from its mirrors in order to
power the injectors. The means by which this is accomplished at high efficiency
is called direct conversion. In the pure fusion case this leads to a large
recirculating power fraction of order unity.

In a simple magnetic mirror (Figure II1I-B-1), as in other containment
devices, the plasma is contained transverse to the axis because of its inability
to diffuse at an appreciable rate across magnetic lines. However, containment
along the axis results from the "mirroring" of individual ion orbits by the
converging field lines at the two ends, where the magnetic field strength B

I11-6
C 3 INJECTOR

r

Ul’

COIL CURRENT

— FIELD LINES

MIN IMUM-B MAGNETIC MIRROR
(YIN-YANG COILS)

FIGURE IIT1-B-1. ITllustrating the Principles of
a Magnetic-Mirror Device in
Minimum-B Geometry

is larger than in the central plane by the ratio R, called the mirror ratio.
An jon (Figure III-B-1) whose motion is directed predominantly toward a mirror
with longitudinal kinetic energy will gain perpendicular (circular) energy Wy
around the field lines as it approaches a mirror. At the mirror it will have
Wy (mirror) = W) (center) x R and will have subtracted correspondingly from
the longitudinal energy. In the case of sufficient W (center) the ions are
brought to rest so that W14 (mirror) = 0. This occurs for those particles

for which w, (center)/w11 (center) is sufficiently large that the direction

of the ion velocity 1lies outside some angle to the axis of the mirror. This

ITI-7
angle defines a corie of directions called the loss cone, such that ions whose
velocity directions 1ie outside it are contained, and the others are lost out
the ends. Collisions between ions can send them into the loss cone and vice
versa. There results a velocity distribution, called a 1oss-cone distribution,
which is not Maxwellian and which largely determines the degree to which loss
may exceed the collisional lower 1imit by influencing the kinds of unstable
plasma waves that may occur.

It has long been known that plasma in the simple mirror geometry is
unstable to magnetohydrodynamics (MHD) motions in which the plasma moves
grossly across the magnetic lines. However, it has been shown theoretic-
ally and experimentally that a system whose magnetic lines are everywhere
convex toward the plasma is stable to MHD modes. Such a system has minimum
field strength B on its axis at the center of the system, and B increases
outward in all directions. The minimum-B system of Figure I1I-B-1 has
fan-shaped ends, one vertical and one horizontal, and the field is supplied
by "Yin-Yang" coils, which are among the most economical of the various
possible coil systems for producing minimum-B mirror fields. This coil

system has been chosen by the Lawrence Livermore Laboratory (LLL) group as
the basis for their reactor design.

To sustain the plasma in a mirror device against collisional end loss it
must be injected with a neutral beam from an injector, as shown in Figure
I11-B-1. The plasma is nearly opaque to this beam and absorbs its energy to
sustain the thermonuclear reactions. The plasma thereby becomes an energy
amplifier because of the total thermonuclear power it produces.

2. Conceptual Engineering Design

The magnetic mirror fusion driver is based upon the Lawrence Livermore
Laboratory (LLL) conceptual mirror-hybrid reactor design.(q) The plasma has
a roughly spherical central portion of radius 2.5 m with mutually perpendicular
"fans" at the ends from which plasma escapes. For the device discussed here

the central ion density n = 9 x 10'7 m™3
20

with B8 = 0.7 and confinement corres-
ponding to nt = 2 x 10 sec/mB. The mean injection energy of the D-T ions
is 125 keV. The neutron wall loading in the first wall is 1.6 Mw/mz.

ITI-8
The magnetic field is furnished by superconducting U-shaped Yin-Yang
magnetic coils of 11-m radius. The maximum magnetic field at the conductor
is 8T, allowing the use of NbTi superconductor. The field of the lower mirror
js 0.5% less than that of the upper mirror, forcing most of the plasma to
escape from the bottom.

Figure I1I-B-2 shows an overall view of the reactor. Magnet, blanket
and primary heat transfer loops are all within a prestressed concrete reactor
vessel (PCRV) which has two holes for the neutral beams and allows access to
the fuel elements through a hole at the top. The PCRV also serves to restrain
the magnets against their internal magnetic pressure. The fission blanket is
made of 600 helium-cooled modules as shown in Figure III-B-3. A single blanket
module is illustrated in Figure I1I-B-4. The helium coolant flows up through
the tritium-breeding pins, out around the fission pin bundle, back through
them and out the diffuser to the steam generators.

Ninety percent of the plasma flow out of the bottom mirror is direct-
converted with a single stage direct convertor with an effective efficiency
of 50%, while the 10% flow of the upper mirror is thermally converted at an
efficiency of 35%.

The two neutral beam injectors are radiation hardened composites of 216
positive-ion, neutral-beam sources delivering 3000 A of 125 keV D and 189 keV
T. MWith direct conversion of the stray beam the injection efficiency is
ny = 0.55. The plasma Q = 0.63 is stated as the ratio of fusion power (400 MW)
to injected neutrsl power (625 MW).

C. LINEAR THETA PINCH

1. Plasma Physics

Unlike other magnetic confinement systems, the theta pinch is a high-beta
device (B = 1) in which very little penetration of the magnetic field into
the plasma occurs. In the theta pinch the plasma density (m1022 m'3) is also
two to three orders of magnitude larger than in the magnetic mirror and
tokamak, and confinement times are correspondingly shorter. The theta pinch

is inherently a pulsed device because of its impulsive method of heating and

III-9
FIGURE III-B-2. Overall View of the LLL-GA

Mirror Hybrid Reactor(3)

ITI-10
Yin-Yang
Mirror Coil
Beam Port
Beam Port
Shield
Helium Inlet
Helium Exit

Blanket
Module
(=~ 600)

Fiqure III-B-3. Cutaway View of the LI(L-GA

Mirror Hybrid Reactor(3

ITI-11
Flow Distribution Dome\ MHR First Wall Module

Diffusers

, / U,Si Pins
1!

1.17M

0.80M

Hex Flange —\

fi,,/— Lithium Hydride Pins

/— 6 ea TZM Bolts
. 2.5 cm Diameter
Pressure Shell Grapple

Evacuated Double
E : Varian Type Seal
)

Vacuum Sinus

B
¥
rg

: Cool Helium

e : ' A
i)

Inner Assy Restraint

Thread and Grapple Plated Metal “O” Ring

and Grafoil Slip Ring

Diffuser Flow Control Orifice
B
Neutron Shield Spike Thermal
Insulation
Hot Helium

Figure TT1-B<1. Mireor: Hybrid Blanket Hodule’?

IT1-12
its high instantaneous power density. For a typical cycle time t. ~ 10 sec,

the duty factor TB/TC ~ 10'3 results in average power densities agd wall loading
which are about the same as for the other concepts. Total magnetic energies

are of the order of 100 GJ, also comparable to those of the other concepts.
However, this energy is pulsed repetitively in and out of the compression-
confinement coil from an external power supply (typically a superconducting
homopo]ar motor-generator whose rotor stores the energy inertially, converts

it to magnetic energy in the theta-~pinch compression coil, and then recovers

it again as inertial rotor energy with a high efficiency (v90%) characteristic

of rotating electrical machinery.

The basic principles of present day theta-pinch experiments are illus-
trated in Figure III-C-1. Ionized D-T gas is produced inside a single-turn
coil by a high frequency oscillating magnetic field in the axial direction.
Following this, a large current (in the poloidal, or theta, direction) is
suddenly fed to the coil from a capacitor bank. This rapidly fills the coil
with magnetic field parallel to its axis. During the dynamic (or “shock
heating") phase, the surface of the plasma is driven rapidly inward by this
axial field, heating the ions and electrons. Later there is a quiescent
(adiabatic compression) phase after the magnetic field is built up on a much
slower (adiabatic) time scale to a steady value in the coil.

A theta-pinch reactor will be a staged theta pinch, so-called because it
employs separate energy sources for the shock heating and adiabatic compression
stages. The shock heating coil is thin and can be liquid metal cooled. It is
connected to a low energy, high voltage circuit whose energy content is a
minor factor in the overall energy storage system. The energy in the magnetic
compression field, which is preponderant, is furnished by a low voltage multi-
turn coil which produces a more slowly rising magnetic field (following the
shock heating field), appropriate to adiabatic compression of the shock heated
plasmas. Such a coil is economical of joule electrical losses, and leads to
a satisfactory excess of reactor power output (low circulating power fraction).
The compression coil is also of sufficient size to accommodate an inner neutron
moderating or hybrid blanket.

ITI-13
SHOCK HEATING
COIL

COMPRESSION
COIL

SHOCK HEATING EQUILIBRATION
IMPLOSION COMPRESSION
f HEATING
(e
o /
|V
© 4
%‘; TIME —
< LOW ENERGY HIGH ENERGY
HIGH VOLTAGE LOW VOLTAGE
RISETIME RISETIME
0.1 psec =1 m sec

ADIABATIC COMPRESSION

FIGURE ITI-C-1. Illustrating the Principle of a Staged
Theta-Pinch Using Separate Shock-Heating
and Adiabatic Compression Coils

2. Conceptual Engineering Design

a. The LASL Designs

There have been two studies of this concept at Los Alamos and a later one
at the University of Washington. The first was based on a capacitively driven
adiabatic compression system with separate shock heating assumed but not
specified in detail. The second LASL study treated the staged heating coil
and its surrounding multiturn adiabatic compression (ACC). The compression
energy store was a set of homopolar generators. Both coils were inside the
fissile blanket, and the 7 to 8 cm thickness of copper detracted from the

I11-14
breeding and blanket energy multiplication. Confinement, and hence nt and the
Q value were assumed to be limited by streaming of plasma out the ends of the
device, which was one kilometer long. The repetition rate of the 10 ms burn
pulses was adjusted to 2.3 Hz to give a neutron-current wall loading 1 MW/m2.

b. The University of Washington Linear Hybrid Reactor

This design remedied some of the difficulties of the LASL designs by
incorporating the following features: ,

(a) Material end plugs were assumed, thereby reducing the energy loss
problem to that of electron thermal conduction.

(b} A reactor core with the hybrid blanket inside the shock-heating
coil and then adiabatic compression coils was used. The main
features of such a core are shown in Figure III-C-2.

(c) A "nhybrid" magnet was used, in which the normal copper pulsed com-
pression coils were placed inside steady state NbTi superconducting
(S.C.) coils. The 8-T field of the S.C. coils is cancelled by a
negative 8-T pulse from the normal coil, shock heating is applied
at zero field, and the plasma compressed in 5 ms to 16 T to a relatively
Tow temperature to lessen thermal conduction and produce a 3.6 ms
plasma burn. The use of this hybrid magnet principle allows a factor
of four decrease in energy and joule losses of the pulsed magnet.

This University of Washington design has therefore been selected for
consideration in this study.

D. LASER INERTIAL

1. Inertial Fysion Physics

The basicAidea of the inertial confinement is to heat an initially
frozen D-T pellet to ignition by the absorption of pulsed radiation in a time
short compared to the time of the pellet disassembly at the burning
temperature (< 10 keV).

ITI-15
ADIABATIC COMPRESSION COIL

\\\\\\\“\’

v

Y

S

ANRNNANNAN

'NNNRNNNANAN
W
=

=
u\
c
r—
>
-
(@]
>

/

\.

R
AN

Y
TR

AN

\ LR

IMPLOS ION
HEATING COIL

RETURN FLUX PATH

(HIGH-PERMEABILITY)
PLASMA

CHAMBER

SEGMENTED
BLANKET HIGH-VOLTAGE
FEEDPLATES
FOR IMPLOS ION
HEATING COIL

FIGURE III-C-2. Section of the Core of a Linear Fusion Reactor
with the Blanket Inside the Multiturn Compres-
sion Coils and Shock Heating Coils

A requirement for ignition is that the range of the fusion-produced
3.5 MeV alpha particles must be short compared to the radius of the

pellet.

of 10° to 10
4.7 x 10

For these conditions to be met the pellet must be compressed by a factor
N above its normal solid or liquid density (0.2313 g/cm3 or

22 ions/cm3). The burn parameter nt is usually expressed in terms of

the pellet radius R traversed at thermonuclear sound speed, and the mass
density of the pellet. A burnup fraction of 30% corresponds to pR:3q/cm2.

I11-16
A figure of merit for the approach to reactor conditions is the pellet
gain factor:

Q - (thermonuclear energy out) = (laser light energy incident on the pellet)

Provided that the plasma burn can propagate from a small central region, Q
values as small as 100 may lead to practical pure fusion plant efficiencies.

2. Conceptual Engineering Design

There have been two in~depth studies of laser driver hybrids by the
Lawrence Livermore Laboratory (LLL) group with the Bechtel Corporation(S) and
the Westinghouse Corporation. The Westinghouse design operates at a neutron
wall loading of 10 Mw/m2 with a blanket power density of 250 MN/m3 with enriched
(3 to 5% Pu) UC fuel but low fissile production. It is optimized primarily to
produce electric power. We do not consider this design although its wetted
first wall concept is important for high energy pellets whose debris fluence
exceeds the capability of a carbon first wall.

Figure III-D-1 shows the LLL-Bechtel reactor. It has a 10 m diameter
first wall (Figure III-D-2) of nonablating grabhite heid at a steady temperature
of 880 K. The structured pellets produce 100 MJ of fusion energy at a repetition
rate of 6.1 Hz, giving a neutron wall loading of 175 MW/mZ. These values
are averaged over a three full power year (4.28 CY at 70% capacity factor} tuel
handling cycle in which the reactor thermal power PTH is held constant at 4000
MWt as the Pu concentration builds up. Fuel management holds the Pu concen-
tration at about 1%. The laser frequency varies from 8.5 to 5.5 Hz to hold
PTH constant. The first wall, shown in Figure III-D-2A is lithium cooled and
sees 25 MJ (KH)kJ/mZ) per pulse (210 MW max.) in the form of X-rays and pellet
debris and 40 MJ (330 MW max.) from neutrons and gamma rays. The remainder
impinges on the upper and lower lithium blankets.

Figure III-D-2B shows the cylindrical side fission blanket and top and
bottom fusion blankets consisting of 50% enriched lithium, beryllium, stainless
steel and graphite. The fission blanket intercepts 66% of the area available
to the pellets. This blarket and top 1lithium blanket 1ift out together as
indicated in Fiqure III-D-2B.

I11-17
gL-III

Pellet injection Lithium to and from

system first wall and
Lt top blanket
/Fuel removal hatch
Top sodium
Top cover RS plenum
/L_El Sodium
P outlet
{ { A~ =Sodium inlet
Top blanket Y/
e ; MWV 2= = Lithium to and
} I Ei [l from radial and
Fission zone : Yl | First walls bottom blankets
Tritium LF
breeding il Reactor
Zones il Point of implosion support
Laser — | w3 Laser
beams ' _,.-;“l beams
! \L Urani -—— Concrete
ranium supports
“\ blanket PPo
N ' Fuel Radial
{ elements  fithium —Jii
Reactor ) ) ~ L‘ blanket U\\ § . Thermal
shell — Bottom blanket shielding
Bottom L ottom blanke :
sodium [
plenum R 2
. S s Bottom blanket support S~
10 meter diam chamber ™~SSseee 00 _ooo==P Blanket segment
16 meter diam shell ———TTmeeesen Lithium support
20 meters high plenums

FIGURE III-D-1. The LLL-Bechtel 40N0-MWt Laser-Fusion Hybrid Reactor(s)

1 mm Molybdenum-

1 mm Molybdenum
corrugatedx mm Y _\
1 mm Molybdenum
Li Void Li Void Li brazed/to graphite
T“[ J‘H Graphite :Ecm

A. First wall structure

Liat 470°C Li at 320°C

| |

! Top blanket

First wall attached
to top blanket

First wall

@

B. First wall outline

FIGURE III-D-2. First Wall Structure of the LLL-Bechtel
Laser Fusion Hybrid Reactor (5

Four 100 kJ lasers drive the pellets in the equatorial plane of the
reactor. They are assumed to be of an excimer type in which 1.2 MeV electron
beams excite Xe gas whose 170 nm fluorescence radiation dissociates C0Se
(carbonyl selenide) to give 489 nm selenium laser light.

The quality Ny is defined as the overall efficiency of the laser from
the electric line, through the 1.25 MV, 2.3 MJ pulsed power conditioner, the
electron beam (75%), the fluorescer (18%), the laser (25%), and the optics
(60 m focal length f/30) (90%). The produce ot these factors is 3.2%. When
power for laser gas conditioning is taken into account the overall laser
efficiency is 1.17% at 5.5 Hz and 1.5% at 8.5 Hz. Over a fuel handling cycle
the time averaged Ny = 1.33%. The gain of the laser pellet system is assumed
to be Q = 250.

111-19
E.

SECTION IIT REFERENCES

V. L. Teofilo, "THR - A Tokamak Hybrid Reactor," TANSA. 32(27), 1979.

J. M. Dawson, H. P. Furth, and F. H. Tenney, Phys. Rev. Lett. 26:1156,
1971. (Also see H. P. Furth and D. L. Jassby, Phys. Rev. Lett. 32:1176,
1974.)

B. Badger, et al., Tokamak Engineering Test Reactor. UWF DM-191,
University of Wisconsin, Dept. of Nucl. Engr., Madison, WI 53706, June
1977.

D. J. Bender, et al. Reference Design for the Standard Mirror Hybrid
Reactor, UCRL-52478, Lawrence Livermore Laboratory, Livermore, CA, May
1978.

Bechtel Corp., Laser Fusion-Fission Reactor Systems Study. UCRL-13796,
Lawrence Livermore Laboratory, University of California, Livermore, CA,
July 1977.

I111-20
IV. FISSION BLANKETS

A. FUEL FORMS

Satisfactory performance of the fusion-fission hybrid system depends a
great deal on the technological basis supporting the selection of the fission
fuel form. Not only is fuel performance important under operating and accident
conditions but fabrication, reprocessing and ultimate waste disposal technologies
must be available or developed. Generally, the technology base for a fuel form
(oxide, carbide, etc.) is dependent on a specific cladding material, geometrical
form (pins, microspheres, etc.), and coolant. The technological basis for UO2
fuel is Timited to fuel clad in Zircaloy or stainless steel, fabricated in pins
and cooled by water. In assessing the status of technology for the fuel forms
of interest for the Tokamak, Mirror, Laser, and Theta-Pinch hybrid reactors all
of the following considerations must be addressed:

e (Oxide Fuel - The most highly developed fuel form of interest for hybrids

is UO2 clad in stainless steel. All commercial experience has been with
pins assembled into bundles. Irradiation performance of water-cooled
S.S.-clad UO2 fuel is fairly extensive. The Liquid Metal Fast Breeder
leactor (LMFBR) Program is rapidly developing Na-cooled data. The Gas
Cooled Fast Reactor (GCFR) Program proposes to utilize LMFBR technology

and has identified differences that must be resolved. The predictable
performance of ThO2 should also be enhanced with this technological base.
Oxide fuels achieve burnups of 40,000 to 100,000 megawatt days per metric
tonne of heavy metal (MWD/MTHM). The transient performance of oxide fuels
is the subject of considerabie R&D in both the Light Water Reactor (LWR)
Safety Program and the LMFBR Program. Extensive development of analytical
methods for design is an integral part of both these programs. The methods
developed will be of use to hybrid blanket designers for determining the
response of oxide fuels to the pulsed power operation of most fusion drivers.
Current transient experiments indicate that oxide fuels containing‘fission
products can withstand only a few rapid transients before failure due tc
mechanical fatigue. It is anticipated that all solid fuel forms will have
this problem due to retained fission products.

IvV-1
e Metallic Fuel - The irradiation performance of many metallic fuels is very
well understood. Of the many alloys and geometric forms that have been
used in production reactors, test reactors, and others, perhaps the most
applicable to fusion-fission hybrid reactors is the U-Fissium pins used

as EBR II driver fuel. U-Fissium is primarily a U-Mo alloy. The pins are
made up of cast U-Mo sodium bonded to 304 S.S. cladding. Burnups of
10,000 MWd/MTHM are current practice. Maximum fuel-clad temperature of
650°C 1imit the application of this alloy-clad combination with helium
coolant. The design constraints for this fuel are well understood so the
steady state performance can be reliably predicted. No transient experi-
ments have been performed, however, so response to pulsed power operation
is unknown.

e C(Carbide Fuel - Design information exists for carbide fuel in two forms.

Stainless steel clad pins have been studied extensively as advanced fuel
for LMFBR's. Although irradiation performance must yet be verified, ex-
perimental programs have been identified and await operation of the Fast
Flux Test Facility (FFTF) for obtaining extensive irradiation data. Burnups
of 100,000 MWd/MTHM are anticipated for fast reactor carbide fuels. The
higher allowable linear heat rating (35 kW/ft compared to 18 kW/ft for
oxide) will not be fully utilized in a hybrid blanket, so the incentive

for carbide fuel in this form is primarily neutronic (higher atom density
of U or Th). The transient response of this type of fuel is unknown. It
is anticipated, however, that the analytical methods developed from current
oxide fuel tests will form a good basis for predicting carbide fuel pin
performance.

The other geometrical form of carbide fuel is the coated particle
technology developed as part of the High Temperature Gas Cooled Reactor
(HTGR) Program in this country and the gas cooled reactor program in
Germany. The coated particles are TRISO or BISO coated beads 200-500 um
in diameter. The beads are imbedded in either a spherical graphite matrix
(Germany) or mixed with graphite in pellet form and put in channels in a
graphite block (HTGR). Extensive experience in helium cooled systems is
available for estimating irradiation performance. Burnups greater than

IV-2
100,000 MWd/MTHM are achieved. The resulting lattice is relatively low power
density (10 kW/2). The transient response of this fuel form has been studied
extensively as part of the HTGR Safety Program. Therefore, adequate methods
for the preliminary determination of response to pulse power cycles exists.

e Silicide Fuel - Uranium silicide (U3Si) has been proposed in some blankets.

This fuel form was developed as part of the CANDU program at AECL. U351
is a~metallic type fuel form. Irradiation experiments with fuel exposure
to 25,000 MWd/MTHM conducted by AECL show little swelling. It has shown
an ability to handle large step increases in power which is important to
pulsed power operation. Its linear heat rating is 20% better than UO2 at
500°C surface temperature. Maximum fuel temperature must be maintained
below 500°C which may 1imit its application in helium cooled systems.
Compatibility of U351 with liquid metal coolants and high temperature
clad materials is unknown.

e Molten Salts - Molten salts have been proposed for hybrid blanket application

primarily as a means of alleviating fuel movement problems in the complex
geometries and because tritium separation would be relatively easy. The
molten salt reactor experimedt (MSRE) demonstrated the feasibility of the
concept; however, many technological questions remain that require develop-
ment. Molten salt is compatible with stainless steel up to 500°C and

with graphite to 600°C. Above that Hastelloy-N must be used. The nickel

in Hastelloy may produce sufficient He in a 14 MeV neutron field to make
embrittlement a problem. Although a development program has been defined
for molten salt fission reactors, it has not been implemented so the bases
for blanket design and salt processing system are very uncertain.

If the various fuel forms are ranked in order of available technology, the
1ist would be:

Oxide fuel in stainless steel clad pins
Coated particle carbide fuel

U-Mo alloy fuel in stainless steel clad pins
Carbide fuel in stainless steel clad pins
Molten salt fuel

Silicide fuel in pins

Oy O BAWw N -

IV-3
How much the technology base should influence the selection of fuel form,
cladding and coolant is certainly a topic for discussion. However, it would
be expected that designs proposed for near-term application would weigh avail-
able technology heavier than designs proposed for ultimate commercial applica-
tion. Considering the near term application of hybrids, available or newly
developed blanket fuels were selected.

The Once-Through and Pu-Recycle blanket designs have the following fuel
form, cladding and coolant combination:

Fuel - UC in rods
Cladding - 316 SS
Coolant - Helium

There is no basis for accurately predicting the performance of this combination.
The overall performance expected from this blanket is superior enough to out-
weigh the technological uncertainty. The Refresh blanket design has the fuel
form, cladding and coolant combination listed below:

Cladding - 316 SS
Coolant - Helium

The fuel and cladding combination for this blanket are very familiar and have
had extensive use in the LWR industry. The fourth fuel cycle, Pu-Catalyst, has
the following blanket composition:

Fuel - Pu0,/U0, (Convertor Region)
- ThC (Breeding Region)

Cladding - 316 SS

Coolant - Helium

This particular fuel cycle will draw heavily on technology developed in the
LMFBR program.

B. TRITIUM BREEDING MATERIAL CANDIDATES

The 1ithium compound selected as the tritium breeding material must satisfy
several requirements. The tritium breeding compound must possess good neutronic
and irradiation characteristics as well as exhibit qgood chemical stability at
blanket operating temperatures. The lithium compound selected must release

IV-4
tritium at a rate so that the tritium inventory in the blanket modules is not
excessive. Lithium compounds fall into the following classes: Metallic, salts
and ceramics.

e Liquid Lithium - Liquid lithium contained in stainless steel rods could be

a potential tritium breeding candidate. The tritium removal would be
complicated, however, by the high solubility of tritium in lithium.
The blanket module tritium inventory would be very high.

e Metallic Compounds - Metallic compbunds of lithium with Al, Bi, Pb, Si and

Sn may be useful for hybrid blankets. The radiation stability of these
compounds has not been established. Also the metallic compounds show the
appearance of liquid phases at low temperatures as the lithium atoms are
transmuted by nuclear reactions in the blanket.

e Nonmetallic Compounds - The oxide-bearing ceramics have the highest

melting points, except for the carbide. The compound L120 has a high
melting point and a high lithium atom density although its vapor pressure
prohibits its use above ~1400°C. It has a strong affinity for water and
carbon dioxide. The reaction,

L120 + H20 = 2Li0H

has calculated free energy change at 298°K of -22.7 kcal so that the
equilibrium vapor pressure of H,0 at 298°K is ~10"]4 torr. Consequently,
the dry powder would be difficult to fabricate without producing some

LiOH which must be dehydrated at an elevated temperature after assembly.

Lithium oxide compounds with A1203 and SiO2 have much lower affinity
for carbon dioxide and water; consequently, these compounds could be
fabricated in dryboxes. The melting point of its lithium rich compound,
L1A102, has been reported between 1610° and 1700°C. Such determinations
were difficult because of the vaporization of LiZO which began ~1400°C,
and caused a change in the composition of the sample. A eutectic liquid
reported at ~1670°C between the compounds L1'2A102 and L1A1508 would form
as the lithium in the compound LiA]O2 is transformed by the neutron
irradiation. The appearance of this liquid and the vaporization of
Tithia 1imits the usefulness of the compound to <1400°C. The desire

IV-5
to avoid excessive sintering of the ceramic compound, L1A102, limits
its usefulness above ~1300°C.

Lithium ortho-silicate, L145104, and meta-silicate, L125103, are
stable compounds which may be useful. The ortho-silicate has a high
lTithium atom density. The ortho-silicate melts, however, by a reaction
with L120, 1255°C, and the rapid vaporization of lithia at this tempera-
ture has been reported. Also, as the Tithium in the ortho-silicate
transforms as a result of neutron irradiation, a eutectic liquid forms
at 1024°C between the ortho and meta-silicates; consequently, the useful
temperature 1imit of the ortho-silicate is <1000°C.

In addition to the oxide ceramics, the carbide of a metal is often
a stable compound. Lithium forms a single carbide L12C2, which reacts
readily with water to yield acetylene. Although the detailed crystal
structure of this compound has not been reported yet, it probably exists
as a salt in which the carbon atoms form a dimer, similar to CaC2 SO
that it is not a stable high temperature compound.

The 1ithium halide salt, LiF, has a high Tithium atom density but
its relatively low melting and boiling points probably limit its
usefulness. Also, the tritium which is generated in a fluoride salt
would be released as molecular TF which may cuase potentially serious
corrosion problems if released into the helium coolant. Consequently,
a low temperature fused salt mixture would have to be circulated to
external equipment for removal of the TF, as has been proposed pre-
viously.

Lithium hybrid or deuteride have many desirable neutronic charac-
teristics as a potential tritium breeding material or neutron moderators.
Their Tow melting point and high hydrogen pressure pose serious 1imi-
tations on their usefulness, however.

Shown in Table IV-B-1 are some of the thermal and physical characteristics
of potential tritium breeding compounds.

Lithium-oxide was selected as the blanket material for one hybrid
reactor analysis in the assessment paper because it has a high lithium
density and high temperature capability. In addition, natural liquid

IV-6
TABLE IV-B-1. Breeding Compound Characteristics(])

Lithium Melting Neutron Reacts
Density Point Tritium Multiplier Chemically With
(atoms/barn-cm)  (C°)  Retention Needed Stable ~__Air
Liquid Tithium 0.042 180 High Yes Yes Violently
Flibe
(47 LiF 53 BeFZ) 0.014 360 Low No Yes No

Solid compounds:

LiAl 0.027 718 Very low Yes Yes Stowly
LiAlH4 0.041 1625 ? Yes ? (dehydride) ?

LiA]O2 0.023 1700 Very low Yes Yes No

Liasi 0.013 635 ? No ? ?

LiZSiO3 0.034 1204 Very low Yes Yes No

Li451‘04 0.050 1256 Very low Yes Yes Mo

L1'7Pb2 0.083 726 Very low No Yes Slowly
L13N 0.041 800 - ? No ? Stowly (?)
Li3Bi 0.040 1145 ? No Yes Slowly (?)
LiZBeZO2 0.038 1150 ? No Yes No

Li,0 0.082 1700 Very low No No (?) No

L 10H 0.037 471 ? Yes. Yes No

LiH 0.059 68 ? Yes ? (dehydride) No

Tithium is used to cool the inner toroidal shield for the Tokamak Hybrid and
the top and bottom cylindrical regions for the Laser Hybrid.

AN

C. COOLANTS

In assessing the technological bases for coolant selection and performance,
several areas need to be considered:

Status of power conversion system components

Availability of design analysis methods and supportive data bases
Compatibility with fuel form, cladding and structural materials
Compatibility with tritium processing requirements

Knowledge of magnetic field effects

Y O W N -

Ability to predict safety performance

In selecting a blanket coolant, the plant power conversion system must be
considered. The plant efficiency versus peak cycle temperature for both the
conventional steam and gas turbine cycles are shown on Figure IV-C-1. These
curves point out that to maintain blanket structural material temperatures within

IV-7
100

80 —

IDEAL CARNOT
CYCLE \\

60 FIGURE IV-C-1. Thermal Efficiency

of Typical Thermo-Dynamic Cycles as

RAEII?TIXESNTEAOI\;J%LYCLE a Function of Peak Cycle Temperature
/

GAS TURBINE BRAYTON CYCLE
WITH RECUPERATOR

THERMAL EFFICIENCY (%)

20 1

0 | ! 1 |
400 600 800 1000

PEAK CYCLE TEMPERATURE (9C)

currently available technology the conventional steam turbine generator will
be employed. Therefore, whatever coolant is selected, the heat transport
system must be made compatible with ultimate transfer of heat to a modern
steam system.

Coolant compatibility with the fission fuel form and cladding is really
the only difference in selection of blanket coolant for a hybrid as opposed to
a pure fision reactor.

®* Water Coolant - In all the areas of technology previously mentioned, we

know the most about water as a coolant. Extensive R&D in the LWR program
has developed an adequate data base and design methods to predict water-
cooled blanket performance. However, water has not been considered as a

blanket coolant to date because it is nearly impossible to remove tritium
from water. In LWRs, tritium releases outside the plant are controlled
simply by limiting the generation of tritium. Impurities (Li) in the core
are reduced to levels which 1imit the tritium production to amounts that
can be released from the plant.

IV-8
Helium Coolant - the HTGR and German Cooled Gas Reactor programs have devel-

oped and demonstrated helium cooled power conversion system technology.
Helium is compatible with all structural materials with the exception of
refractory metals and alloys. The impurity levels attainable in real
systems result in corrosion problems for the refractories.

To get adequate heat transfer and transport properties, helium systems
have to be operated at relatively high pressures (50 to 70 atms.). In the
complex geometries of hybrid blankets, this results in a requirement for
structural material fractions which increases parasitic absorption of
the neutrons. Where cladding and structural materials are stainless steel,
helium-cooled systems yield 30% power conversion efficiency. If higher
temperature alloys (TZM, Inconel, etc.) are used, efficiencies approaching
40% are projected. Helium has good neutronic properties with no
anticipated MHD or corrosion enhancement effects in magnetic fields.

Liquid Metal Coolants - The LMFBR program is developing data and system

components for Na cooled systems. The major uncertainties in Na cooled
systems are the MHD effects in rapidly changing high magnetic fields and
the effects of magnetic fields on corrosion and mass transport rates. Due
to enhanced heat transfer, higher sodium temperatures can be achieved with
stainless steel structural materials and thus power conversion efficiencies
near 40% can be achieved without the use of high temperature alloys. The
LMFBR Program is also developing an extensive data base for Na coolant.
These data will be directly applicable to assessing hybrid performance.

The use of liquid Li as a coolant has not been investigated for a
hybrid blanket. Although it is attractive neutronically for producing
tritium, the technology base for Li is uncertain. Li appears to be more
corrosive than Na and hence operating temperatures must be lower (50°C)
to be compatible with stainless steel, resulting in Tower power con-
version efficiency. The increased corrosion and mass transport rates
result in uncertainty in the applicability of current Na power conversion
system components.

IvV-9
Because liquid metals can be used at low pressures, they result in
low structural material requirements. Where magnetic field effects are
not important (vertical confinement applications) designers have proposed
using both Na and Li as coolants, thus maximizing the use of R&D benefits
from the LMFBR Program.

If candidate coolants are ranked by the available technology base, they
would fall in the following order:

1. Water coolant

2. Helium coolant
3. Sodium coolant
4. Lithium coolant

The blanket coolant selected for this study is helium because it is
unaffected by magnetic fields, and because it is compatible with tritium
breeding and recovery concepts.

D. HEAT TRANSFER - FLUID FLOW

The four hybrid blankets, in general, do not pose serious heat transfer-
fluid flow design problems compared to fission reactor technology. A good
measure of this is the relative power density in hybrid blankets compared to
various fission reactor cores as shown in Table IV-D-1. The fuel-coolant
lattices being selected by designers are typical of GCFR and LMFBR technology;

hence, there appears to be some freedom in increasing the amount of fuel in
the blanket.

TABLE IV-D-1. Typical Reactor Power Densities

PWR  GCFR  LMFBR  HTGR  Hybrid

Average Core Power 100 240 360 8 20
Density
Maximum Core Power 285 360 540 13 100

Density (MW/m3)

The calculational methods for heat transfer and fluid flow, developed by the
fission reactor programs, are adequate for conceptual hybrid reactor blanket de-
signs. However, detailed design and safety analyses of start-up and pulsed opera-
tion are going to require much closer coupling of thermal and mechanical analysis
methods than now exist for both fuel and structures.

Iv-10
E. STRUCTURAL DESIGN

In assessing the status of structural design of fusion-fission hybrid
blankets, three areas must be addressed:

e Materials properties
e Structural layout
® Design analysis

The comments here pertain to hybrid blanket structure. The magnet shield region
also has important structural implications; however, hybrid designers are currently
relying on the pure fusion reactor blanket and shield program to develop the

shield requirements because of the much lower neutron flux and energy entering

the shield region for the hybrid.

e Materials Properties

A1l components of a fusion-fission hybrid blanket are subjected to
large fluences of high energy neutrons (>1 MeV). When selecting materials
and projecting performance, irradiated materials properties are important.
The most complete irradiated properties data currently comes from the LMFBR
Program which has concentrated on the 300 series stainless steels. The
LMFBR Program ranges from extensive theoretical studies of damage mechanisms
to establishing the bulk properties necessary for the designer. Data and
correlations exist or are being developed for swelling and helium embrittiement
due to irradiation. Irradiated stress rupture and cyclic fatigue data also
exist. Stainless steel is serviceable up to 600°C with sodium or helium
coolant, somewhat Tower for Tithium or molten salts (500°C). For conceptual
designers to change to alternate cladding and structural material to achieve
higher operating temperatures would introduce a great deal of uncertainty
into predicting design adequacy and structure lifetime.

The adequacy of the LMFBR data to predicting performance in a high
14 MeV neutron flux is of concern to designers. The current OFE materials
program, however, is running some preliminary experiments to see if irradi-
ation damage (swelling and helium embrittliement) are different for 14 MeV
neutrons than LMFBR correlations predict. These experiments along with
LMFBR data will form the only firm design bases available until high energy

IV-11
neutron test facilities are in operation. Extensive materials properties
data will not be available on alternate materials before the time frame of
interest for initial hybrid operation (1990-2000).

e Structural Layout

Structural layout of current fusion-fission hybrid designs depends a
great deal on the geometry of the fusion driver. Figure IV-E+l is a modu-
lar arrangement developed by PNL in this study for the Tokamak Hybrid.

In the tokamak modular concept, the fuel pins are oriented radially.
The helium coolant enters from the supply header, flows along the outer

module wall, turns 180° and flows back through the fuel region to the
coolant exit header (see Figure IV-E-2). In some vacuum system concepts,
the vacuum seal is formed where the modules connect to the header. In
others such aS the one whoen, a separate vacuum barrier is designed. A
separate vacuum barrier (first wall) simplifies module design since the
high heat 1oads from plasma losses are taken by a separate structure.
There are 11 madules located around the torus segment (see Figure IV-E-3).
The neutral beam injection port occupies 10-15% of the first wall space and
will extend completely around the torus. The torus will be divided into
60 segments each having 11 blanket modules to make a total of 660 modules,
A close-up view of a Tokamak Hybrid module is shown in Figure IV-E-4, The
thermal or mechanical stresses in the stainless steel module wall due to
the 700 psia helium coolant pressure will be well below the maximum allow-
able 50 ksi provided the walls are externally supported and/or they have a
double wall construction,

The Mirror Hybrid utilizes a cylindrical module design shown in
Figure IV-E-5. These modules are arranged in orange peel shaped segments
(Figure IV-E-6). There are approximately 600 modules arranged into 16
reactor segments. Figure IV-E-6 shows the overal segment placement around
the plasma chamber. In the cylindrical module design, coolant gas enters
through the inlet duct and fills the plenum below the fertile fuel rods.
The gas then passes through the space provided between the submodule's
side walls and the blanket fuel rods. At the first wall, the flow is

IV-12
! /
/
/
/
- - /
FIRST
WALL T
POLOIDAL
FIELD COIL
SHIELD~—}” 7 fr, %
BEAM
PORT —

) I

FIGURE IV-E-1. Tokamak Hybrid Blanket Segment

IV-13
-4
BEAM PORT

jew of the Tokamak

Cross Section V
id Reactor

Hybr

FIGURE IV-E-3.

IV-15
91-Al

HELIUM
MAN [FOLD

s )

MODULE BODY 0 \-N-/ 2N
%0 'e.'(/// ::;% UL

‘ FERTILE
FUEL PINS

FIGURE IV-E-4. Tokamak Hybrid Module Detail

~—— FIRST WALL

<«—— FERTILE FUEL
~1.0 :
\ /
-+
} f ~—Li,0 PINS
- \ CIL
3 o GRAPHITE REFLECTOR
- 4l
\ \
|
— \
e
X

\

FIGURE IV-E-5. Mirror Hybrid Blanket Submodu1e(2)

IV-17
SELECTIVE LEAKAGE PORTS

il
[

COILS

BLANKET
SEGMENTS

FIGURE IV-E-6. Mirror Hybrid Blanket Arrangement(2)

Iv-18
reversed and directed into the blanket region by flow baffles. The
helium then passes through the fission zone and tritium breeding zone
and is discharged through a duct into a main manifold pipe.

The Laser inertial hybrid blanket arrangement is shown in
Figure IV-E-7. It is a segmented type of blanket structure and utilizes
extended modular fuel assemblies similar to the ones designed for the
Tokamak Hybrid.

The Theta-Pinch hybrid is a linear device composed of 200 blanket
modules. The total length is 500 meters with each module being 2.5 meters
Tong. Figure IV-E-8 shows a schematic drawing of the module and fuel pin
arrangement.

F. MECHANICAL AND THERMAL HYDRAULIC DATA

For the purpose of this hybrid assessment study the fissionable and
tritium breeding fuel assemblies for the blanket modules for all drivers
were assumed to be similar to the Tokamak hybrid blanket module assemblies.
This allowed the neutronic calculations performed for the Tokamak Hybrid
(see Section V) to be scaled for all corresponding drivers with appropriate
factors for fusion power and blanket coverage. The corresponding mechanical
and thermal hydraulic information for these combinations of driver with
blanket-fuel cycle options are tabulated in Tables IV-F-1 through IV-F-4.

In all cases the coolant flow rates and velocities are adjusted to obtain
the corresponding outlet/inlet temperatures. At helium inlet pressure of 700
psia, this corresponds to velocities in the range of 10 to 100 m/s with an

approximate heat transfer coefficient of 1 to 2 w/cm2°C.

IV-19
| LITHIUM

PELLET INJECTION TO AND FROM
SYSTEM FIRST WALL AND
COVER _—TOP BLANKET

....... Pesanasnia,,,,

| , FUEL REMOVAL HATCH
TOP BLANKET
SUPPORT TOP SODIUM

PLENUM

TOP COVER

SODIUM
OUTLET

= SODIUM INLET 9 I’

LITHIUM

FIRST WALLS TO AND FROM
- -F LLS - . RADIAL AND
FISSION ZONE " \BOTTOM BLANKETS
4 .
TRITIUM B \
BREEDING | it \
ZONES - / E ] e \
, |~ REACTOR
: B SUPPORT
ELLL PO _ R LASER
3 m S — —
| I T POINT OF IMPLOSION — BEAMS
‘ | " -)=—~ CONCRETE
| R URANIUM SUPPORTS
Dot BLANKET —
: I { |
: .= 1 |FUEL
: T RADIAL
3 o
: | } {' ELEMENTS LITHIUM
! T BLANKET — j
REACTOR 4 b— — THERMAL
SHELL - C , /_/ SHIELDING
BOTTOM - N |
SODIUM e '
PLENUM LLBOTTOM BLANKET i

| BLANKET SEGMENT
BOTTOM BLANKE T SUPPORT SUPPORT

PLENUMS

10METER OitAM CHARAIBER
16 METER DIAM. SHELL
20 METERS HIGH

SODIUM REMOVABLE CAP FOR
OUTLET PIPE FUEL ELEMENT CHANGE

SO0IUM
INLET.PIPE

LITHIUM INLET
{TOBOTTOM
LITHIUM PLENUMI

UTHIUM OUTLET S ]

{FROM TOP LITHIUM
PLENUM)

3SETS OF FUEL ELEMENTS r
BOUNDED BY §.S. SHEATHS.
EACH SET HAS 3 ROWS OF

19 ROD CLUSTER URANIUM
FUEL ELEMENTS IN 27 HEX [
AGONAL TUBES PER ROW
OR 81 FUEL ELEMENTS PER SET HH -

]

RADIAL LITHIUM BLANKET BOTTOM SODIUM PLENUM
S.S. BODY C/W LITHIUM DOUBLE WALLED
COOLANT PASSAGES AND

$.S. CLAD CARBON

FIGURE IV<E-~7. Laser Hybrid Blanket Segment Arrangement(S)

IV-20
L2-A1

Multi region blonket
Neutron multipker (zone 1)
Enriched seed (zone 2- 4}
Refleclor (zone 5)

Qutlet qgas coolant
manifold -

- ~
\.v\i AN
Gas coolant

Concrete biclogical
shield

Raits for removal of

Steel hoer for——" o bottom half of blanket

vacuum chamber .

Current -

feed
plates(2) -

)

. \\
To vocuuin

FIGURE IV-E-8.

High- pressure envelope

=

To staging and carnpression
. ECCID(X:'IR)( banks

Implosion heating -~ -
PFN copacitors -7 -
- - ’

Inlet gas coolant
- manfold

Wn‘)s \‘\\\' ) . g o .
~ e
\\ /,/
\c

Linear Theta-Pinch Hybrid Blanket Module
TABLE IV-F-1. Tokamak Hybrid Mechanical and Thermal

Hydraulic Information

Reactor Perameter

Reactor Thermal Eower (thh)
Fusion Power (thh)
Electrical Output (Mwe)Net
Core Design:

Blanket Heat Qutput:

Fission (thh)

Li Reactions (thh)
Specific Power (thh/MT)(a)
Power Density (N/cm3)(b)

Geometric Information:
Fast Fission Zone Height {(cm)
Number of Blanket Modules
Fuel Pins (Rods)/Module

L120 Pins/Module

Overall Module Dimensions (LxWxH)cm

Module Material
Cladding Parameters:
Fue1/L120Rod:

Outgide Diameter {cm)
Wall Thickness {mils)
Cladding Material

Fuel Type

Blanket Coolant -

Outlet/Inlet Temperature (°F)

Pu- Pu-
Recycle Catalyst
4,144 6,603
1,160 1,160
1,000 1,835
2,615 5,136

252 191

8.6 16.5

54 68

26 39

660 660
2,500 2,500
600-800 600-800
84x40x78 24x40x78
S.S. S.S.
1.0/2.0 1.0/2.0

15 15
S.S. S.S.
uc U02/Pu02 ThC
Helium Helium
1200/600 1200/609

{a) Based on blanket fiscion power and total fuel loading.

Refresh
3,715
1,160

853

26
660
2,500
600-800
84<40x73

S.S.

1.0/2.0

S.S.
UO2
Helium

1200/600

{b) Pcwer density determined from blanket fission power and fuel reaion volume.
TABLE IV«F-2. Mirror Hybrid Mechanical and Thermal
" Hydraulic Information

Reactor Perameter ReE;;1e CaéZ#Qst Refresh
Reactor Thermal Power (thh) 2,578 3,603 2,404
Fusion Power (thh) 402 402 402
Electrical Output (M‘.-Je)Net 139 544 71
Core Design:
Blanket Heat Output:
Fission (thh) 1,082 2,125 915
Li Reactions (Hwth) 46 28 40
Specific Power (M, /MT) ) 3.75 . 6 3.9
Power Density (W/cm3)(b) 20 25 16.8
Geometric Information:
Fast Fission Zone Height (cm) 26 39 26
Number of Blanket Modules 580-600 580-600 580-600
Fuel Pins (Rods)/Module 2,200 ~2,200 2,200
LiZO Pins/Module 500-600 500-600 500-600
Overall Module Dimensions (m) g$;32§er 0.8-1.0 0.8-1.0 0.8-1.0
Hodule Material S.S. S.S. S.S.
Cladding Parameters:
Fue]/LiZORod:
Outside Diameter (cm) 1.0/2.0 1.0/2.0 1.0/2.0
Wall Thickness (mils) | 15 15 15
Cladding Material S.S. S.S. S.S.
Fuel Type uc UOZ/PuO2 ThC UO2
Blanket Coolant - Helium Helium Helium
Outlet/Inlet Temperature (°C) 530/280 530/280 530/280

(a) Based on blanket fission power and total fuel loading.
(b) Pcwer density determined from blanket fission power and fucl reaion volume,

IV-23
TABLE IV-F-3,

Reactor Pecrameter

Linear Theta-Pinch Mechanical and Thermal

Hydraulic Information

Reactor Thermal Power (My

Fusion Power (thh)

Electrical Output (Mwe)

Core Design:

Net

Blanket Heat Qutput:

Fission (thh)

Li Reactions (MW
Specific Power (thh/MT)(a
Power Density (H/cm3)(b)
Geometric Information:

Fast Fission Zone Height (cm)

Number of Blanket Modules

th)

Fuel Pins (Rods}/Module

Li,0 Pins/Module

2

Overall Module Length (m)

Moaule Material
Cladding Parameters:

Fue]/LiZORod:

th)

Qutside Diameter (cm)

Wall Thickness (mils)

Cladding Material

Fuel Type

Blanket Coolant -

Outlet/Inlet Temperature (°C)

(a) Based on blanket fission power and *ntal fuel

)

Pu- Pu-
Pecycle Catalyst Refresh
4,835 8,197 4,343
1,098 1,098 1,098
45 1,557 -176
3,477 6,829 2,940
150 92 127
1.4 2.4 1.7
8.7 10.1 7.4
26 39 26
200 200 200
5000-6000  5000-6000  5000-6000
2000-3000  2000-3000  2000-3000
2.5 2.5 2.5
S.S. S.S. 5.5.
1.0/2.0 1.0/2.0 1.0/2.0
15 15 15
5.S. S.S. S.S.
uc UOZ/PuO2 ThC UG,
Helium Helium Helium
850/540 850/540 850/540
1oading.

(b) Pcwer density determined from blanket fission power and fuel reaion volur:.

IV-24
TABLE IV-F-4. Laser Hybrid Mechanical and Thermal
Hydraulic Information

Reactor Pzrameter

Reactor Thermal Power.(thh)
Fusion Power (thh)
Electrical OQutput (MNe)Net
Core Design:

Blanket Heat Output:

Fission (thh)

Li Reactions (thh)

Specific Power (thh/MT)(a)

Power Density (H/cm3)(b)
Geometric Information:

Fast Fission Zone Height (cm)

Number of Blanket Segments

Fuel Pins (Elements/Segment)

Overall Segment Height (m)

Module Material
Cladding Parameters:
Fue]/LiZORod:

Outs%de Diameter (cm)
Wall Thickness (mils)
Cladding Material

Fuel Type

Blanket Coolant -

Outlet/Inlet Temperature (°C)

(a) Based on blanket fission power and total fuel

Pu-
Recycle

3,300
850
940

10

S.S.

1.0/2.0
15

S.S.

uc

Helium

470/320

loading.

Pu-
Catalyst

4,980
850
1,567

S.S.

1.0/2.

15
S.S.

UOZ/PUO2 ThC
Helium

470/320

Refresh
3,015
850
830

S.S.

Helium

470/ 320

(b) Pcwer density determined from blanket fission power and fuel region volume.

IV-25
G. REMOTE DISASSEMBLY AND MAINTENANCE

The blanket lifetime for the four hybrid fuel cycles is on the order of
four years. The radioactivity and decay heat levels resulting from the hybrid
blanket operation necessitates a blanket designed for remote maintenance. The
blanket module arrangement must also provide ease of access and disassembly.

A cross section view of the Tokamak is shown in Figure IV-G-1, In order
to gain access to the blanket modules, the foltowing operations have to be
performed:

1. The upper and lower blanket shield and VF coil have to be raised.
2. The hinged shield would then be swung open and secured.

3. Helium supply and return lines would then be disconnected from the
module manifolds.

4. The welds and seals adjoining adjacent segments would then be cut
or machined off.

5. Finally, the blanket segment would be either lifted out of the
reactor by an overhead crane or transferred by means of a remotely
operated carriage.

The blanket segments are transferred to a hot cell operations area. Here
the segment would be remotely dismantled and the fuel rods removed from each
individual module. The fuel rods would then be placed into special canisters
and retired to a decay heat spent fuel storage basin. The L120 pins and
reflector region would be placed back into the module along with fresh fuel
pins. Then the segment would be reassembled and placed into the reactor.

The Mirror Hybrid Reactor blanket maintenance strategy is influenced by
the large size of the blanket sections. Figure IV-G-2 shows the blanket
module concept used in the Mirror Hybrid Design. Each segment is a separate
pressure vessel which makes vacuum leak testing an easier task. During a
blanket replacement outage, one-fourth of the segments are replaced. The
steps that would be necessary for blanket access are listed below.

IV-26
LZ-A1

PARTICLE

COILS

A
/‘/é/
",'
,L,

HITTIIIY
I

/L

U

S il S B N 5N S N

%

gl SUPPORT

/' STRUCTURE

COLLECTOR
PLATES OVERHEAD SUPPORT
7 STRUCTURE
e TFCOIL
4 7
g N
N
Z ) \
W~ 777 Do VF COILS
RN, NEUTRAL BEAM
WA X7 e SHIELD INJECTOR
| o
| | s L WSS S A VI A VLS SIS/,
e SEPARATRIX
* B y A
' ’\\/ A
7~ PLASMA ,
Al |
N |~ © N neurraL beam”
» N A VACUUM PUMPS
N 4y - .
Al CRYO-SORPT [ON
PANELS
CHEVRON
_ SHIELD
5

a

FIGURE IV-G-1. Tokamak Hybrid Reactor Cross Section (%’

4.
5.

4
ZE:;E%TT:\::§§7q‘I!!|l""iiiiiiiiiiiiiiiiiiifiii‘

Yin-yang coil

FIGURE IV-G=2. Mirror Hybrid Reactor Blanket
Module(2
Remove beam injectors by means of the overhead crane.

Disconnect the top vacuum shell (Fiqure IV-G-3) and remove to
another location.

Hoist the removable top plugs and transfer to a temporary storage
location. |

Hoist the upper magnet and transfer it to a holding area.

Disconnect the shield dome thus exposing the blanket segments.

The blanket segments are hoisted by means of the overhead crane and

transferred to a hot cell workshop. Here the helium ducts can be disconnected

making the segments easier to manipulate. The spent fuel is stored in a decay

IvV-28
Load-bearing
/" insulation

~End tank
; PCRV

-~ {_Vacuum
—| shell
+ Top
magnet
Magnet and blanket | " . |_Load-bearing
support structure | % 3 % : ‘ — | insulation
Shield———————— /7 (I8 ' [@E&N—— . - Bottom
PR ARy a ! | ) " . | maanet
Blanket — Tt J* x

- -~ - -l ) a
» L ¥ ¥ 7

FIGURE IV-G-3. Mirror Hybrid Reactor Bl?nket
and Structure Components(2)

heat removal basin. Fresh fuel is loaded into the individual submodules and
then the assembled segment is placed back into the reactor.

An alternate blanket maintenance approach is shown in Figures IV-G-4 and
IV-G-5. In this method the blanket submodules are individually arranged
forming a spherical plasma cavity. The technique for blanket replacement in
this configuration is to use an in-chamber removal and replacement method.

Blanket replacement for the Laser Hybrid and Theta-Pinch Hybrid will

also be performed by remotely operated fuel handling machines. For the Laser
Hybrid one of the eight segments is removed by means of an overhead hoist.
The segment itself can be further disassembled into three sets of fuel elements.
The fuel elements are manipulated by a grapple and hoist crane similar to the
fuel transfer machine used in LWRs, Blanket access in the Theta-Pinch Hybrid
is accomplished by removing a portion of the high pressure shell that encom-
passes the blanket fuel rods.

IV-29
e . e\
L i H e o Hi F—{
??
' i (fl
T ] '
3 4
A DN\ e .
] — n
[_—Jlf—gj
\Q <
P | ?
‘ A [ il
] ]
5 6
A5 B\ A5 N\
= T3 - =

}“— =
e L)

FIGURE IV-G-4. Alternate Blanket R?p1acement Technique
for Mirror Hybrid(2

IV-30
MOTORIZED
UPPER SUPPORT

SUPPORT

COLUMN
PIVOTING,

TELESCOPING
IVHM ARM

N.  DIRECT
« CONVERTER 11t
~

e

FIGURE IV-G-5. Mirror Hybrid Module Handling Machine(2)

Iv-31
H.

SECTION IV REFERENCES

Conceptual Design Study of a Non-Circular Tokamak Demonstration Fusion

Power Reactor. GA-AL3992, General Atomic Co., San Diego, CA, November

1976.

D. J. Bender, et al., Reference Design for the Standard Mirror Hybrid
Reactor. UCRL-52478, General Atomic Co. and Lawrence Livermore

Laboratory, Livermore, CA, May 1978.

Laser Fusion-Fission Reactor Systems Study. Bechtel Corporation and

Lawrence Livermore Laboratory, Livermore, CA, July 1977.

B. Badger, et al., Tokamak Engineering Test Reactor. UWFDM-191,
University of Wisconsin, Dept. of Nucl. Engr., Madison, WI 53706, June
1977.

IvV-32
V. NEUTRONICS

A. COMPUTATIONAL METHODOLOGY

The primary objectives of the neutronics calculations were to determine
the fissile fuel and power production in various fission blankets combined
with various fusion drivers. Neutronics computations were performed for all
selected blankets adapted to the Tokamak Hybrid Reactor. The results of such
computations were then appropriately scaled to obtain the neutronics perform-
ance data for the selected blankets combined with the three fusion drivers.
The calculational model for the Tokamak hybrid reactor is shown in Figure V-A-1.
In order to represent the various blanket types, the materials of Zones 15,
16 and 17 were changed in each of the calculations. The remainder of the
reactor remained the same.

(1)

The neutron flux calculations were made with the computer code ANISN
which solved numerically the one dimensional Boltzman equation. The geo-
metrical model was in vertical cylindrical geometry and is identical to that
described in Figure V-A-1. Reflective left hand and vacuum right hand bound-
ary conditions were employed. A neutron source that varied both in space
and energy was used in the calculation. A 58-P3 numerical solution was used.

The cross sections for the transport calculations were generated from
ENDF/B-1IV fi]es(z) into a thirty energy group structure. The methodology is
discussed under Nuclear Data.

The burnup calculations were made with the code ORIGEN;(3) It is a point
code and uses one group average cross sections to determine the isotopic
contents of the fissile and fertile nucleus as a function of operating time.
The ORIGEN Tibrary did not include cross sections to 14 MeV, thus it is neces-
sary to generate the cross sections for input into this code. This is accom-
plished in the following manner.

. f n-2n  _n-3n . .
Four cross sections, o , cc, c ., 0 , ‘are required for each isotope.

An ANISN calculation is made for a particular fuel, for a particular time in
the 1ife of the blanket segment and the calculations began with the blanket
segment at the beginning of life. Those isotopes which are not present at
this time, such as the higher isotopes of Pu are placed in the calculation
at a low concentration. This does not affect the numerical calculation of

V-1
“L=¥Y-A FNIDIA

J13eWdYdS [euoL3e|[nNd|e) 403003y

x
—
O QA
3 -
e
wr ~
o
3
(1]
c I———————~
| <
(@]
=
230 ————- -
(]
S
310 ~ SRR
313 |[Z2-_337
[Rag=
o LV —
- 50°
/3
o T-Breeding 95 Li, 5 SS
358 | ————— -
o Reflector
90 C, 5 SS, 5 Li
378 | ——
| T-Breeding
95 Li, 5SS
428 | = -
o SS MWall
429.5 | ———m -
w C Liner
430 —_
< Vacuum
440 |—
i =
) oY)
[ g
I &
|
640 ———
~ Vacuum
659.5
o C Liner
660 —m——
= SS Wall
665 -
50° Fertile & = 10 SS
o Fissile Fuel . 40 He
678
50" Fertile & : 10 SS
o Fissile Fuel ' 40 He
691 [ =
T-Breeding
3 707 L120. 10 SS, 20 He
708 |— -—
T-Breeding
= 70" Li20, 10 SS, 20 He
7 eee—
| Reflector
‘2 807 C, 107 SS, 10 e
730 —- -—

| T-Breeding
|8 707 Li20, 10° S, 20 He
743 S
the flux, but does allow for the calculation of a reaction rate from which a

particular cross section is obtained.

The reaction rate for a particular isotope, reaction and zone is found

from the following relationship using calculated fluxes.

Rn = T Y 0.. V onk oy
= ) .. V. o, .
k jin ¢ i 0o
where
n . . . .
Rkis the reaction rate for reaction n for isotope k in zone ¢,

¢1j is the group energy flux in energy group i and interval j,

yjis the volume of interval j,

o?k is the cross section for reaction n for element k and energy
group i, and

Ply is the number density of element k in zone &,

The average cross section, o, for input into ORIGEN is then calculated
from the relationship:

— n
o R, /[ = re.. V.]p
k kg iing 1 1] J kg

and the average flux @ is determined by

¢ = [ = v e.:. V.l & V.
jin 2 i jing
. . ¢ f n-2n
In the calculations here, the average cross sections for ¢, o , ©
and Gn—3n for the following isotopes were generated:

241, 242 244

Pu, Pu, and Cm.

Since in a fusion reactor, the flux and fission density changes rapidly

V-3
in the blanket, ORIGEN calculations are made for different zones in the
fertile and fissile blanket. This involves the calculation of the average
cross sections and flux for each zone.

Based on a one year burnup calculation isotopic generation and depletion
are determined for each zone. These isotopic concentrations are then input
into ANISN for new flux calculations and the process is repeated. The process
may be repeated for as many years as desired. The accuracy may be improved
by decreasing the zone width, in effect creating more zones for which cal-
culations are made and decreasing the time period for the burnup calculation,
as the flux is assumed constant during this period of time.

B. NUCLEAR DATA

The cross sections for the transport calculation were generated from
ENDF/B-IV into thirty energy groups and covers the energy range from 18 MeV
to thermal. ETOG(a) generates the epi-thermal and fast data and FLANGE(S)
is used to process the thermal data. These group cross sections are pro-
cessed with scattering matrices expanded in Legendre polynomials. All the
cross sections generated in this manner are infinite dilute, and thus the
important isotopes, such as U?35, U238 and Th?32 must be resonance self-
shielded.

The shielded cross sections are generated with the cell code EGGNIT(S).
In this code a typical unit cell in the fissionable lattice is mocked up.
A unit cell calculation is made using a fine energy group structure to deter-
mine the flux shape in the unit cell. The Nordheim Integral Treatment is
used for resonance self-shielding. The resultant cross sections are then
flux weighted and resonance self-shielded.

Unfortunately, the cross sections in the EGGNIT library extend only to
10 MeV and contain only a P; expansion of the scattering matrix. Thus it
is necessary to substitute the EGGNIT generated microscopic absorption and
fission cross section into the ETOG generated set and re-normalize the total

232Th, 235U, and 238

cross section. Three isotopes U are treated in this

manner.

For isotopes in which the thermal cross sections play an important

role, such as 6L1, 235U and 238Pu, the thermal cell code GRANIT(7) is

V-4
used to generate the thermal cross section and these cross sections replace
the FLANGE generated thermal cross section.

C. FISSILE FUEL BREEDING

Three separate blanket types were considered:

Pu02-U02 CONVERTER followed by ThC2 Zone

In Figures V-C-2, 3, and 4 the outside blanket is represented for each
of the three types. The different zones correspond to Zones 15 to 20 in
Figure V-A-1.

In the first case, the fuel is natural uranium in the form of UO2 in
a 26 cm thick zone. In Case 2, the fuel is natural uranium in the form of
UC in a 26 cm thick zone. In Case 3, natural uranium in the form of UO2 is

mixed with 232Py0

9 in an equilibrium mixture in a 13 cm thick zone. This
is followed by a ThC2 zone, 26 cm thick. The volume fraction of all fuels
is 50% with 10% stainless steel as the structural material. Helium is the
coolant and occupies 40% of the volume.

ANISN calculations were made to determine the neutron flux from which

239Pu and 233U production rates were calculated, The

the fission rate and
results for the three blankets at the initial start up are given in

Table V-C-1.

Comparing the UO2 blanket with the UC blanket, it is obvious that UC
is the best blanket, both from the standpoint of the number of fissions per
fusion and the Pu production. This arises from the fact that although both
blankets have the same volume percent of fuel, the UC blanket has more U
per cm3 than the UO2 blanket. Thus, more U is exposed to the 14 MeV flux
than for the UO2 blanket, which results in a greater number of fast fissions
and more secondary neutrons. The greater the number of secondary neutrons,
the greater the number of neutrons being absorbed in Pu and tritium breeding
reactions.

V-5
9-A

| Fertile & Fissile Tritium Breeding Tritium
Fuel Zones Zones Reflector Breeding
& pi4— 48 P4 > 14 -) i ¢ 2
50% UO2 50% UO2 70% LiZO 70% L120 80% C 70% LiZO
10% SS 10% SS 10% SS 10% SS 10% SS 10% SS
40% He 40% He 20% He 20% He 10% He 20% He
Nat. U - Nat. U - Enriched Enriched Enriched
0.7% 233y | 0.7 23% | 90% OLi 90% OLi 90% OLi
Zone 15 16 17 18 19 20
Radius 665 678 691 704 717 730 743
(cm)
FIGURE V-C-2. U0, Blanket Schematic

LA

Zone

Radius
(cm)

Fertile & Fissile

Tritium Breeding

Tritium

Fuel Zones Zones Reflector Breeding
¢ D4 pie >4 »e >4 +»
50% UC 50% UC 70% L120 70% Li,0 80% C 70% Li,0
10% SS 10% SS 10% SS 10% SS 10% SS 10% SS
40% He 40% He 20% He 20% He 10% He 20% He
Nat. U Nat.U23 Enriched Enriched - Enriched
0.7% 2| 0.72 0| 904 6L | 907 OLi 90% 6L i
15 16 17 18 19 20
665 678 691 704 7117 730 743

FIGURE V-C-3.

UC Blanket Schematic
8-A

Radius (cm)

Convertor
Fissile Fertile Fuel Zone Tritium Reflector Tritium
Fuel Breeding Breedin
4 D¢ pl¢ Pl ———> ¢ - ¢ 1)
50% - 50% ThC 50% ThC 70% Li,0 |80% C 70% Li.0
00~ - PuO 2 2 2 2
2 2 |10% s.5s. 10% S.S. 10% S.S. 10% S.S. 10% S.S.
10% S.S. 40% He 40% He 20% He 20% He
40% He
7.2% 23%, Enriched Enriched
0.7% 235, Li Li
92.1% 238 902 oL 9020 ;
665 678 691 704 717 730
15 16 17 18 19

FIGURE V-C-4.

PuQ2-U02~ThCo> Blanket Schematic

743

20
TABLE V-C-1. Blanket Neutronic Characteristics

Net Fissions Net Pu Net 233U Tritium

Per Source Production per Production per Breeding

Blanket Neutron Source Neutron Source Neutron Ratio
UO2 0.186 0.440 - -1.17
uc 0.220 0.618 - 1.19
Pu02-U02 0.424 0.047 1.24 1.18

ThC2
239

The third case in which a converter of UO2 in
ThC2 blanket combines the best features of both fuel cycles. The high
fission cross section for 14 MeV neutrons and resulting ~5 neutrons per
fission coupled with the high thermal absorption in Th for 233U production
results in the best blanket from the standpoint of fissile fuel production.
An equilibrium mixture of Pu02 results in a larger power output, due to thermal
fission, compared to cases one and two which had only 0.7% 235y.

PuO2 is followed by a

D.  TRITIUM BREEDING

In each case following the fertile and fissile zones are the tritium
breeding regions. These regions contain stainless steel clad pins of L120
enriched to 90% in 6L1 to capitalize on the thermal flux. Few fast neutrons
remain this far from the fast neutron source and Tittle tritium is bred by
the 7Li(n, n-a)T reaction. Most of the tritium is bred by the 6Li(n,a)T
reaction. For this reason it is essential in a helium cooled Tokamak to
employ enrichments in 6Li in the tritium breeding compound. The use of
natural Li coolant may obviate this need in Tokamaks and other confinement
geometries. In fact, natural Li compounds could be used in this Tokamak
geometry with helium coolant to obtain tritium breeding ratios >1 but only
at the expense of fissile fuel production.

In the inside blanket, natural 1iquid lithium is used. Here, advan-
tage may be taken of the fast flux for the breeding of tritium without the
loss of a neutron in the 7Li(n, n“a)T reaction. However, here also the
bulk of the tritium is bred through the 6Li(n,a)T reaction.

V-9
The results for the three blankets at the initial start up are given
in Table V-C-1.

In each case the tritium breeding ratio is greater than one with suffi-
cient excess that each could be self sustaining for tritium requirements. If
tritium had not been bred, the neutrons could be used for fissile fuel pro-
duction with the increase in production rates of fissile fuel being greater

per source neutron than the number of tritium atoms produced per source
neutron.

E. BURNUP AND ISOTOPICS

The burnup calculations which determine the isotopic buildup and
depletion are sensitive to the average flux used in the calculation. This
is due to the fact that 239Pu and other fissile isotopes which build up
greatly affect the average power determined by ORIGEN.

15

For example, an average flux of 1.21 x 10 n/cmz-sec for a five

year period of time produces an average power of 15.51 MW per metric ton
of U. An average flux of 2.91 x 10]5 n/cm2-sec over a five year period
produces an average power of 45,15 MW per metric ton of U, Both cases had

the same initial conditions of natural UO2 as fuel,

The average flux is determined by the fusion power level and the
isotopic concentration used in the ANISN calculation. One year burnup
was chosen as the time period for which the flux should remain almost
constant. The isotopic concentration of major isotopes following a one
year burn is shown in Table V-E-1 for Zone 15 and 16 of Figure V-C-1.
This is a UO2 blanket containing natural uranium as initial conditions.
The Table 1ists only the isotopes that would be used in the next ANISN
calculation,

Note that the isotopic concentration buildup in Zone 15 is much
larger than that in Zone 16. The 14 MeV flux and fast flux in Zone 15
is much larger than in Zone 16. The fast flux is responsible for many
of the reactions. Therefore, the average cross in Zone 16 is smaller
than Zone 15. The average flux in Zone 16 is smaller than in Zone 15.
As expected much of the Pu build will occur in the first few centimeters
until it reaches an equilibrium concentration of about 8%.

F. FISSILE FUEL AND POWER PRODUCTION

The fissile fuel and fission power production in the hybrid blanket
combined with the Tokamak Fusion driver have been calculated from the
blanket neutronic characteristics as computed by the ANISN and ORIGEN codes
as displayed in Tables V-C-1 and V-E-1. The annual fissile fuel pro-
duction as well as the thermal power averaged over the four year fuel
management cycle (see Section VII) are tabulated in Table V-F-1. The
plant availability (0.75) as well as the driver duty factor which is in-
herent in the calculated fusion power (see Section III) have been taken in-
to account in computing these averages, in addition to the effective
blanket coverage for the penetration of the beam parts, divertor channels,
and poloidal field coils.

The corresponding amount of fissile fuel production and average blanket
fission thermal powers for the other fusion driver systems with the selected
blankets have been computed by scaling from the Tokamak results and they are
also listed in Table V-F-1. For each driver blanket combination the corre-
sponding differential in fusion neutron power duty factor and blanket coverage
were taken into account in determining the average values. It should also be
noted that in addition to the blanket fission power, the thermal power includes
the fusion neutron power distributed in the blanket and shield as well as the
fusion alpha power and any radiation generated in the fusion plasma incident
on the first wall.

Due to the similarities in first wall thickness and blanket coverage, the
scaling of the Tokamak neutronic results agrees favorably with the laser hybrid
calculations of others. However, due to significant dissimilarities of the
same nature, with the other drivers, this scaling becomes somewhat pessimistic
for the mirror and theta pinch hybrids.

The heating rates for the three blankets combined with the Tokamak driver
are shown in Figure V-F-1. These rates are calculated for the reactor at start-
up. For the UC and the UO2 blanket the power production is due almost entirely

V-11
TABLE V-E-1. Isotopic Concentrations After One Year Operation

Gram-Atom per Metric Tonne of Uranium

Zone 15 & 16
Initial Zone 15 Zone 16
Isotope Concentration ’One Year One Year
Th '2.75 x 1078 8.1 x 107°
Th 5.8 x 1078 1.8 x 1078
Pa | 2.7 x 1078 2.8 x 1078
Pa 2.6 x 107 3.5 x 10712
U 1.5 x 107 1.3 x 107°
! 1.4 x 1073 3.7 x 107
U 1.9 x 1072 5.8 x 107°
U 30.2 26.7 28.4
U 3.99 1.22
u 4170 4100 4140
y 2.5 x 1073 1.41 x 1073
Np 7.51 2.26
Py 1.77 x 1073 1.53 x 107
P 1.07 x 1072 1.49 x 1073
Py 37.5 21.1
Py 0.584 0.16
Py 8.9 x 107 1.08 x 1073
P 5.2 % 107° 3.24 x 1070
Cm 8.9 x 1070 1.32 x 10711
TABLE V-F-1. Blanket Fissile Fuel and Fission Power Production

uo, uc PuOZ-UOZ/ThC2
Driver/Blanket Fuel kg Pu/yr  Mit kg Pu/yr Mt kg 233U/yr MWt
Tokamak 1388 2439 1950 2867 3810 5327
Mirror 574 965 807 1128 1575 2153
Linear Theta Pinch 1845 3067 2592 3627 5066 6421
Laser Inertial 941 2150 1323 2450 2584 4130

to the fast fissions. The power profile in these two blankets almost parallels
the fast flux in the blankets shown in Figure V-F-2. Both the power and fast
flux are down by two orders of magnitude after transversing the 26 cm fuel
region of the blanket. In these blankets little may be gained in the way of
power by increasing the thickness of the fission blanket.

In comparing the UO2 with the UC blanket, it may be noted that the UC
blanket produces the greater amount of power. In the UC blanket, uranium
occupies a greater percent of the volume of the zone than in the UO2 blanket,
even though the volume percent of the fuel (UO2 or UC) is the same in both
cases, i.e., 50%. This increased density of uranium results in a greater per-
cent of the uranium being exposed to the fast flux resulting in a greater
number of fissions. Because of the higher density the fast flux decreases
faster in the UC blanket than in the UO2 blanket as noted in Figure V-F-2.
Thus increased power production may be obtained by increasing the ratio of
ZfiSSiO”/Z?BZ?rgtlggucture in the blanket. For example, U-Mo could be expected
to be a better fuel than UC in terms of power production.

The most dramatic changes in power production may be obtained by enriching
the fuel. Fiqure V-F-1 compares the heating rates of U02-Pu02 blanket with the
UC and UO2 blanket. The U0, and the UC blanket contains natural uranium, while

2
the Pu0,-U0, blanket contains 7.2% 239PuO. The power profile is not as steep

2 2
in the mixed oxide blanket indicating a considerable amount of power is being

generated by the fission of 239Pu, rather than by fast fission of U. This is
200

100
70

50
40

30

20

10

7.0

HEATING RATE, Wicm>

5.0
4.0

3.0

2.0

1.0

0.5

FIGURE

@puoz—uo2
@uc
- ®uo,
- (:)ThC2
i HEATING RATES AS
s A FUNCTION OF
RADIUS FOR THREE
i BLANKET TYPES
L —
[
=
- zZ
<L
x
L. (@]
[—
pd
2 P
o o
(@]
(& ]
o
zZ
<
- z
- > = -
I~ —
- w
o
- | i | L 1
660 665 670 680 690 700 710

RADIUS, c¢cm

V-F-1. Heating Rates as a Function of Radius

for Three Blanket Types

v-14

® - GROUP FLUX PER SOURCE NEUTRON

1x107

1x107°

10

1x10°

by

FIRST WALL

MeV FLUX

FERTILE AND
FISSHLE BLANKET

14,19T014,92___

l TRITIUM

TRITIUM 1
BREEDING

1REFLECTOR l BREEDING

—_ UO2 BLANKET

~——— UC BLANKET

A FAST AND THERMAL
GROUP FLUX AS A
FUNCTION OF RADIUS

\
)
THERMAL
FLUX |
\
N
N
N
N
N
| I | | I L 1 | _
660 670 680 690 700 710 7120 130 140
RAD{US, cm '
FIGURE V-F-2. A Fast and Thermal Group Flux as

a Function of Radius

also evident from Figure V-F-3 which shows the perturbation in the thermal
flux in the converter region.

The fast flux in all three cases has been degraded by several orders of
magnitude by the time it reaches the tritium breeding zones as shown in
Figures V-F-2 and V-F-3. Thus, little tritium from the 7Li(n,na)T reaction
will be produced. These zones have been enriched in 6L1 to increase the
macroscopic cross section for the 6Li(n,a)T reaction and therefore reduce the
parasitic absorptions which do not result in a tritium atom.

Fast fission will occur in thorium. However, the cross section is small
compared to the fast fission cross section for uraniuf. Thus for any reason-
able power production a uranium converter is needed. At 680 cm in Figure V-F-1,
the fast flux is about the same for the U02, UC and ThC2 blanket, yet the power

production in the ThC2 is about 7 w/cm3 while the U02 and UC blankets are

about 35 W/cms.

It is important to note there is no fissionable material in the inside
blanket (Zone 7). It was felt that the area was inaccessible and only liquid
lithium was placed there as it could be pumped. Had the reactor been large
enough that a uranium converter could be placed in this zone a significant
increase in both fuel and power production could be obtained. In the current
configuration a large percent of the fusion neutrons do not enter a uranium
containing zone, and therefore do not have the opportunity to cause a fast
fission.

In conc]usion,‘increased performance in blankets may be obtained by
increasing the fusion neutron power and the ratio of the macroscopic fast
fission cross section to the macroscopic absorption. The most dramatic
increase in power is obtained by increasing the enrichment in the blanket.
Although ThC2 gives impressive fissile fuel production it requires a uranium
converter, otherwise a serious decrease in performance occurs.

V-16
® - GROUP FLUX PER SOURCE NEUTRON

1x1074

1x10™

1x107°

1x10°

10

Pu0,-UO TRITIUM TRITIUM
2 eral } et ¢ RERECTOR
l lCONVERTOR ThC, BLANKET BREEDING REFLECTOR BREEDING

FIRST WALL

14.197014.92_—~
MeV FLUX

A FAST AND THERMAL
GROUP FLUX AS A
FUNCTION OF REACTOR
RADIUS

THERMAL FLUX

l

660 670 680 690 700 710 7120 130 140
RADIUS, cm

FIGURE V-F-3. A Fast and Thermal Group Flux as a
Function of Reactor Radius

SECTION V REFERENCES

W. W. Engle, Jr., A Users Manual for ANISN. K-1693, Oak Ridge National
Laboratory, Oak Ridge, TN, 1967.

H. C. Honeck, ENDF/B, Specifications for an Evaluated Nuclear Data File
for Reactor Applications. BNWL-50066, Brookhaven National Laboratory,
Upton, NY, 1966.

M. J. Bell, ORIGEN - The ORNL Isotope Generation and Depletion Code.
Oak Ridge National Laboratory, Oak Ridge, TN, May 1973.

D. E. Kusner, et al., ETOG-I, A Fortran IV Program to Process Data from
the ENDF/B File to the MUFT, GAM, and ANISN Formats. WCAP-3845-1, ENDF-114,
Westinghouse Electric Corporation, December 1969.

H. C. Honeck and D. R. Finch, FLANGE II, A Code to Process Thermal Neutron
Data from an ENDF/B Tape. DP-1278, Savannah River Laboratory, Aiken, SC,
1971.

C. R. Richey, EGGNIT: A Multigroup Cross Section Code. BNWL-1230,
Pacific Northwest Laboratories, Richland, WA, November 1967.

C. L. Bennett, GRANIT: A Code for Calculating Position Dependent Thermal
Neutron Spectra in Doubly Heterogeneous Systems by the Integral Transport
Method. BNWL-1634, Pacific Northwest Laboratories, Richland, WA, November
1971.

VI. CONCEPTUAL PLANT DESIGN

This study has concentrated on coupling the fusion driver (tokamak, mirror,
laser and theta-pinch) and the blanket fuel cycle (once-through, Pu-recycle,
Pu-catalyst and refresh). Therefore, only minimal effort has gone into assess-
ing and characterizing conceptual plant designs. The following discussion
deals with the plant layout, energy conversion system, and primary system
vessel and piping for the four hybrid reactor concepts.

A. PLANT LAYOUT

1. Tokamak Hybrid Reactor

The fusion driver being used is a scale-up of the University of Wisconsin
TETR. The blanket module design was discussed in previous sections of this
report. A schematic diagram of the reactor hall and its dimensions is given

in Figure VI-A-1.

The power conversion system is shown in Figure VI-A-2. The thermal
storage system provides two functions. First, it combines the energy collected
from all sources, i.e., inner shield, divertors and blankets, into the main
helium circuit. Secondly, it minimizes temperature fluctuations in the helium
coolant due to the pulsed operation of the tokamak driver.

2. Mirror Hybrid Reactor

The conceptual plant layout for the mirror reactor is given in
Figure VI-1-3. The containment structure has a 96 meter diameter and is
approximately 75 meters high.

The mirror hybrid coolant system is composed of a primary and an auxiliary
loop. Normal blanket coolant during reactor operation is provided by the
primary cooling loop. Emergency shutdown cooling is provided by the auxiliary
system. The location of steam generators and the arrangement of the primary loop
containment structures are shown in Fiqure VI-A-4. Also shown is the helium
ducting and manifold system for the blanket module seaments.
65 m

k

IR NN

)

50 m

Y

*

=
—t

O
D N -

— — — — — S— — VPR S—— ——— —

— e — ——— —— — ———  — — ———

— — — —— — — — — — —

Tokamak Hybrid Reactor Hall

-A-1.

FIGURE VI

VI-2
E-IA

—yL "
T
3 2

System r Steam Reheat

é } Wet Cooling
inrer ‘/ Gerlerotor Tower

Steam Steam r[j:]

Blanket

Ny
|

Compressor

] Tritium 1

Recovery
Plant

I Preheaters

Tritium

FIGURE VI-A-2. Power Conversion System for Tokamak Hybrid Reactor

Turbine
/ hail

7

Containment

New blanket
assembly

Administrative
and auxiliary

Large component
fabrication and
maintenance

|

Hot blanket operations

96 mdiam—— o

“Q

Plant layout — plan view

/Containmem

\ Polar crane

Large component
fabrication and

N\

Plant layout — elevation

FIGURE VI-A-3. Mirror Hybrid Plant La_yout(”

VIi-4
CHECK VALVES AT OUTLET (8 EA)

STEAM GENERATORS {12 EA)
TURBO-

SURPENTINE
MANIFOLD

HOT DUCTS
(12 EA}

TRIM VALVES
AT OUTLETS
(12 EA)}

(a) Vessel Arrangement

INJECTOR ENVELOPE

TOROIDAL
MANIFOLDS

(b) Blanket Manifolding

1
FIGURE VI-A-4, Mirror Hybrid Reactor( )

VI-5

CIRCULATORS
Cooled helium Teaves the steam-generator through flow control valves and
enters the circulators. The trim valves control circulator power requirements
between loops. From the steam-driven, turbo-circulator outlets the coolant is
transferred through check valves to a large diameter ring manifold surrounding
the reactor itself. Coolant is bled from the ring manifold through 50 radial
ducts containing blanket-flow control valves; the valves match the coolant flow
to the heat load of the blanket modules. Helium then moves through the vertical
distribution channels within the blanket structure, through the modules, and
outward into the manifold. The heated coolant is directed toward the 12 steam
generators to complete the primary power conversion loop circuit. An attractive
feature of the PCRV (Prestressed Concrete Reactor Vessel) is that the primary
power conversion loop helium never leaves the concrete structure, which gives
the primary loop great integrity and renders a sudden depressurization impossible.
The permanent blanket structure is also embedded in the PCRY, which eliminates
a maintainable interface between the power conversion loop and the blanket.

A1l power conversion loop maintenance is done either on the inner face of the
reactor sphere by replacing blanket modules remotely or on equipment lTocated in
top-head cavities with standard gas-cooled reactor technology. The contain-
ment building crane is used to remove and replace top-head cavity closures and
power conversion l1oop equipment within the cavities, including neutral beam
injectors, circulators, and steam generators.

The primary coolant Toop system consists of 50 12-module blanket units
which are connected by ducts and manifolds. There are 12 steam generators,
eight helium circulators, five auxiliary heat exchangers, and five auxiliary
heljum circulators. Fiqgure VI-A-5 shows the primary helium loop schematic
as well as the secondary coolant loop arrangement. These were taken from
Reference 2.

3. Laser Hybrid Reactor

The reactor building layout is shown in Figure VI-A-6. The laser hybrid
reactor building is approximately 52 meters high (from ground level) and, 1like
the mirror hybrid facility, it has an overhead crane system used for blanket
maintenance.

VI-6
L-IA

AUXILIARY
BLANKET LOOP

MAIN BLANKET LOOP

FLOW
CONTROL
VALVES._|

COLOMANWOLD//
(1.83)

1
\k\\\\\\HOTMANWOLD
(1.98)
V v,
A A {0.76)
FEN
AN N

MAIN
END-TANK
LooP

(0.66)}—

BLANKET
QUADRANTS

{0.76) —

(1.58)

STEAM

GENERATORS
- 3Im0Ox 11.6mH
{f - 3.7mDx 13.9mH

TRIM VALVES -
HELIUM AUX
CIRCULATORS CIRC
7 - 15,1 MW 5
1 - 111 MW 4
5§ — 2.7MN 8
AUX. CIRC.
CHECK CHECK
VALVES __. VALVE ] I ]
(1.53)

N COLD MANIFOLD (1.83)
NOTE: DUCT DIAMETERS SHOWN (N METERS{ )}

FIGURE VI-A-5a. Mirror Hybrid Coolant Systems (Primary Coolant Loop

8-IA

ol AESUPERA-
"l HEATER

g EQUIVALENT LIVE STEAM LOSSES

_@_. TO MELIUM CIRCULATOR BEARING
WATEA PUMP TURBINES

Ha CIRCULATOR

SUPER-
HEATER
S
EVAPO-
RATOR
—_—
ECOMO-
MIZER

1

LP TURBINE

()

GLAND
STEAM
CDNDENSER

PRESSURE

AEGULATOR

MAKE - UP

TURBINE
HP TURABINE
STEAM
GENERATOR
b {
B/F
TURBINE
Y. y 4 y \ 4
WP e HP P P LP
HTR || HTA (ag— HTA o HTA HTR HTR
ND 3 ND.2 NO ! NO. ] NO 2 NO 1
b 4 4
4 4 4 Py 4
PUMP
FROM INJECTOR
WASTE HEAT FAOM HELIUM v
EXCHANGER CIRCULATOR REARING @

FIGURE VI-A-5b.

WATER PUMP TURBINES

Mirror Hybrid Coolant Systems (Secondary Coolant System

(1)

TD INJECTOR
WASTE HEAT
EXCHANGER
— e ¥82m
EMERGENCY

COOLING

STACKS

|
M —

N ) S
SECONDARY
COOLANT
LOOP PUMP

e

STEAM

GENERATORS
¥ ::%A___. i
‘gz}f§9'J

AEACTOR CELL

+31m

STEAM
GENERATOR
BUILDINGS

GRADE = Om

LASER TUNNELS

SECONDARY
SODIUM
STORAGE

_—t3m

SECONDARY

T2 RECOVERY l ‘ REACTOR CELL EQUIPMENT SODIUM LOOP i
EQAMIPMENT — -1 \\ TRITIUM . —2Im
RECOVERY
NEQUIPMENT

FIGURE VI-A-6. Laser Hybrid Reactor Building Layout(z)

4. Linear Theta-Pinch Hybrid Reactor

Details of the reactor hall for the theta-pinch hybrid are given in
Figure VI-A-7. The reactor building is approximately 500 meters long and
9.5 meters high. Running the length of the reactor is an overhead module
handling crane. Beneath the reactor is situated a rail carriage for blanket
removal from the bottom portion of the hybrid.

B. POWER ANALYSIS

A schematic diagram for the Tokamak Hybrid Reactor is shown in
Figure VI-B-1. Accompanying plant parameters for each of the fusion
blankets are listed in Table VI-B-1. The neutral beam injectors supply
200 MW to the plasma during the three-second startup phase of the
reactor cycle. An additional 405 MW is required for the remainder
of the reactor support systems. A breakdown of the THR recirculating

VI-9
OL-IA

s S

DI SCONNECT

VACUUM LINE AND\

HIGH PRESSURE 5

HELIUM GAS OUTLET,

|——1.5 m —

15 TON CRANE FOR UPPER
/mmn REMOVAL

OIL STRUCTURAL STEERL
SUPPORT

HIGH PRESSURE HELIUM
GAS INLET

INLET MANIFOLD

? ELECTRICAL DISCONNECTS

—T0 PFN AND

POWER SUPPLY

=o}

R an b
e |
] ]
i §
’ e
— ‘a‘
F S 5 55 5, a
! - - 3
QUTLET | /
MANI FOLD ; *\ /
\\ ) ’/
"

FIGURE VI-A-7.

8LOCK S

/RAIL CAR FOR LOWER BLANKET REMOVAL

le——————PRIMARY CONTAINMENT AND

RADIOACTIVE SHIELD

/Pl PE CLAMP AND FLANGE \l\fl

COMPRESSION FIELD

MODWLE SUPPORT -

J S
L0

HELIUM GAS COOLANT MANIFOLDS

Linear Theta-Pinch Hybrid Reactor Configuration
LL-IA

410 MW RECIRCULATING POWER

t
405
Mw SYSTEM
SUPPORT
200 1160
Mw | NEUTRAL TOKAMAK \ mw | BLANKET/SHIELD COnERMAL | GE
EeTo v RS AL e EFFICIENCY
FOR |INJECTORS AND DIVERTOR -
3 SEC , =34

FF

PLUTONIUM-239
OR
URANIUM-233

FIGURE VI-B-1. Tokamak Hybrid Plant Schematic

NE
SL-IA

TABLE VI-B-1%

Tokamak Hybrid Plant Parameters

Thermal Power Total Thermal Plant Gross Net Fissile Fuel
Due tc Fissions Power Efficiency Electric Electric Production Rate
Blanket FP (MW) TP (MWt) E GE (MWe) NE (MuWe) FF (ka/yr)
Pu-Recycle/Once Through 2615 4150 0.71 1410 1000 1950 Pu
ThC,-Pu Catalyst 5135 6600 0.83 2445 1835 3810 1233
2210 3715 0.675 1263 853 1390 Pu

UO2 - Refresh

*See Figure VI~B-1.
power requirements is given in Table III-A-3. The ignited plasma
supplies 1160 MW of fusion power to the blanket, shield, first wall and
divertor.

The Theta-Pinch Hybrid Reactor is depicted in Figure VI-B-2 with
various plant parameters for each of the three fission blankets listed
in Table VI-B-2. The compression energy source for the theta-pinch
draws an enormous 4050 MW of electric power, of which 2025 MW is
delivered to the plasma and 1920 MW is directly recoverable. The over-
all efficiency of the compression energy system is 95%, requiring only
2130 MW of recirculating power. The 500 meter plasma delivers 1098 MW
of fusion power to the blanket, shield and first wall.

The Laser Hybrid Power Plant is shown in Figure VI-B-3. .0f the 300
MW recirculating power required, 225 MW is needed to run the laser system.
Operating at an efficiency of 1.5%, each of the four lasers delivers 100 kJ
at a frequency of 8.5 Hz. With a pellet gain of 250, this corresponds to
a fusion power of 850 MW. The tritium breeding blankets on the top and
bottom of the reactor chamber receive 34% of the fusion power and supply
600 MW of thermal power to the turbine system. The remaining 64% of the
fusion power is delivered to the radial fission blanket, shield and first
wall. The performance of each of the three blankets studied is indicated
by the parameters listed in Table VI-B-3,.

The Mirror Hybrid System shown in Figure VI-B-4 develops 404 MW of
fusion power with the driving neutral beam drawing 1094 MW. The particles
streamin¢c out of the fan ports are gquided to the direct energy conversion
system which supplies 377 MW thermal power to the turbine system and produces
234 MW of electric output. Eighty percent of fusion power is delivered to the
fission blanket and shield. The resulting plant parameters for the Mirror
Hybrid System are listed in Table VI-B-4.

VI-13
L-IA

2130 MW REC IRCULATING POWER

FIGURE VI-B-2.

.- —e
; 1920 MW RECOVERABLE
COMPRESSION POWER
BLANKET/ THERMAL
4050 mw | COMPRESSION 15095 M /" THETA P SHIELD E
w4050 MW HETA PINCH CONVERTER
ENERGY SOURCE FUSION POWER AND EFFICIENCY]
FIRSTWALL 7 -.45

FF

PLUTONIUM-239
OR
URANIUM-233

Theta-Pitch Hybrid Plant Schematic

NE
SL-1A

TABLE VI-B-2%*

Theta-Pinch Hybrid Plant Parameters

Thermal Power Total Thermal Plant Gross Net Fissile Fuel
Due tc Fissions Power Efficiency Electric Electric Production Rate
Blanket FP (MW) TP (MWt) E GE (MWe) NE (MWe) FF (ka/yr)
Pu-Recycle/Once Through 3480 4835 0.0207 2175 45 2590 Pu
ThC,-Pu Catalyst 6830 8200 0.423 3690 1560 5070 %33
2950 4350 - 1950 ~-175 1845 Pu

UO2 - Refresh

*See Figure VI-B-2.
91-IA

300 MW RECIRCULATING POWER

FIGURE VI-B-3.

0 MW

Laser Hybrid Plant Schematic

RADIAL
FISSION
BLANKET/SHIELD

- <
ISMW [ SYSTEM
SUPPORT TRITIUM BREED ING
200 MW .| TOP AND BOTTOM | 600 MW
P BLANKETS/SHIELD
AND FIRSTWALL

ELLET THERMAL
{ 225 mw LASTEER Rippin CONVERTER
% SyoTEM EFFIC IENCY

FF

PLUTONIUM-239
OR
URANIUM-233

7-.375

LI-TA

TABLE VI-B-3¥ Laser Hybrid PTant Parameters

Thermal Power Total Thermal Plant Gross Net Fissile Fuel
Due tc Fissions Power Efficiency Electric Electric Production Rate
Blanket FP (MW) TP (MWt) E GE (MWe) NE (MWe) FF (ka/yr)
Pu-Recycle/Once Through 1775 3300 0.758 - 1240 940 1325 Pu
ThC,-Pu Catalyst 3485 4980 0.839 1870 1570 2585 1233
UO2 - Refresh 1500 3015 0.735 1130 830 940 Pu

*See Figure VI-B-3.
8L-IA

1122 MW RECIRCULATING POWER

POWER

FIGURE VI-B-4.

- <
S -
DIRECT ENERGY e
CONVERTER
D-T
PLAS MA 7
| 1094 N\w NEUTRAL BEAM SuapoanmvNG l t
INJECTORS A -

CONVERTER

EFFIC IENCY
FISSION 7=.39
BLANKET —
FF

PLUTONIUM-239
OR
URANIUM-233

Mirror Hybrid Plant Schematic

6L~IA

TABLE VI-B-4* Mirror Hybrid Plant Parameters

Thermal Power Total Thermal Plant Gross Net Fissile Fuel
Due tc Fissions Power Efficiency Electric Electric Production Rate
Blanket FP (MW) TP (MWt) E GE (MWe) NE (MWe) FF (ka/yr)
Pu-Recycle/Once Through 1080 2580 0.111 1260 140 810 Pu
ThC,-Pu Catalyst 2125 3600 0.327 1665 545 1575 0233
UO2 - Refresh 915 2400 0.059 1195 70 575 Pu

*See Figure VI-B-4.
C. SECTION VI REFERENCES

1. D. J. Bender, et al., Reference Design for the Standard Mirror Hybrid
Reactor. UCRL-52478, General Atomic Co. and Lawrence Livermore
Laboratory, Livermore, CA, May 1978.

2. Laser Fusion Fission Reactor Systems Study. Bechtel Corporation and
Lawrence Livermore Laboratory, University of California, Livermore, CA,
July 1977.

VI-20
VII. HYBRID FUEL CYCLE ANALYSIS

The fuel cycle options for the four fusion drivers will be characterized
- s0 that a nonproliferation assessment of these systems can be made. Chapter
VII wiil be divided into three parts: A. Fuel Alternatives, B. Fuel Manage-
ment Studies, and C. Facility Requirements.

In Section A, the fueling alternatives section, two scenarios will be
discussed: 1) no-reprocessing, and 2) reprocessing to recover fissile material
for recycle purposes. The relationship between the hybrid and LWR reactor will
be developed for the reprocessing scenarios.

Fuel management strategies for the tokamak, mirror, laser and theta-pinch
hybrid reactors will be discussed in Section B of this chapter. Blanket
management information such as module Tifetime, maximum exposure, blanket
replacement time, and the number of modules replaced each year will be identi-
fied for each hybrid device. The initial quantities of fertile fuel, stainless
steel structure, L120 tritium breeding material, and graphite reflector will
be determined. In addition, the 30-year blanket fuel charge and discharge
amounts for some of the fuel cycles will be determined.

In Section C, facility requirements, the blanket fuel rod fabrication
and module reprocessing facilities will be described. This description will
facilitate the identification of potential diversion points and possible pro-
liferation paths.

A. FUELING ALTERNATIVES

In order for facility descriptions to be developed, a characterization
of each fuel cycle is needed. There are four fuel cycle scenarios being
investigated for the hybrid reactor:

1. Once-Through - natural uranium fueled hybrid in throwaway mode (power
production only).

2. Pu-Recycle to Thermal Reactors - hybrid with dual role of fissile fuel
production and power production.

VII-1
3. Refresh Fuel Cycle - hybrid reactor re-enriching PWR fuel and returning
re-enriched fuel to PWR. Another Refresh type fuel cycle being inves-
tigated is a denatured U-235 (20%) in U-238 [PWR-U5(DE)/U/Th]. The
spent PWR fuel from the cycle, about 9% U-235/U will be re-enriched in
the hybrid with the buildup of the fissile isotope U-233.

4. Pu-Th (Pu Catalyst) Fuel Cycle - hybrid reactor breeds U-233 in a
plutonium-thorium target; U-233 is then recycled in LWRs while the
plutonium is recycled in the hybrid.

The prolifieration resistance which may be attributed to a reactor and
its associated fuel cycles may only be estimated by assessing the facility
and speed with which weapons usable material can be extracted or diverted
from the systems involved. These systems may be "hardened" to resist
extraction or diversion by technical and institutional measures. In prin-
ciple, any such measures or "fixes" which may be available to fission
reactor fuel cycles can also be employed in the hybrid reactor system.
Moreover, hybrid systems may have some unique nonproliferation advantages
over fission breeder reactors since the spectrum of their copious source of
fusion neutrons may be tailored in appropriate blanket designs which are
more readily adaptable to such technical fixes.

In order to place the hybrid concept in some nonproliferation per-
spective it may prove useful to relate the candidate hybrid fuel cycles to
the fuel cycle scenarios and technical fixes being considered for fission
reactors. Such a perspective may give some indication as to whether these
fuel cycles possess desirable nonproliferation qualities which may permit
the appropriate criteria for proliferation resistance to be achieved.(]) We
consider two scenarios: (1) no reprocessing of spent fuel, and (2) repro-
cessing of spent fuel to recover and recycle fissile materials in fission
reactors. In the reprocessing scenario we examine recovery and recycle of

233

(a) denatured U in fission reactors, and (b) plutonium in fission

reactors.

1. No-Reprocessing

The Light Water Reactor (LWR) is the main type of commerical reactor
in operation in the U.S. The High Temperature Gas Cooled Reactor (HTGR)

VII-2
has seen limited commercial deployment. The LWR fuel cycle in the case of no-
reprocessing in which spent fuel assemblies are stored is outlined below and
the hybrid fuel cycle options related to these.

The current once-through LWR fuel cycle is shown in Figure VII-A-1. The
spent LWR fuel is shown going to storage where it stays until such time that a
decision is made as to its ultimate disposition.

In the no-reprocessing scenario, the hybrid role is limited to producing
power for sale. The hybrid fuel cycle analogous to the once-through LWR cycle
is shown in Figure VII-A-2. Natural uranium in the form of uranium carbide is
used as blanket material for the hybrid. The blanket is irradiated, the uranium
fissions, and power is generated. The spent blanket is discharged and tempo-
rarily stored in a decay heat removal area similar to LWR spent fuel pools
awaiting ultimate disposition.

Another hybrid fuel cycle that operated in the no-reprocessing mode is
the "refresh cycle." This is shown in Figure VII-A-3. Natural uranium is
mined and refined in order to produce uranium dioxide for fabricating blanket
modules. The blanket is irradiated in the hybrid where neutrons are captured
in U-238 to produce Pu-239. The bred Pu blanket material is inserted in a
fission reactor to produce power. After the fuel is depleted in the fission
reactor it is sent back to the hybrid to be "refreshed" in Pu-239. Upon
refreshing, the fuel is again used in the fission reactor for power production,
Fuel might be shuffled between fission reactor and hybrid two to three times
depending on the obtainable fuel 1ife, After this cycle the spent fuel is
stored for ultimate disposition. It is possible that the fuel would reguire
refabrication between irradiations to remove the bulk of the fission products
and extend fuel life.

In addition to the "refresh" cycle discussed above, the hybrid reactor
might be used to "refresh" or "re-enrich" normal spent fuel (i.e., as in
Figure VII-A-1 where the fresh fuel is enriched to 3% 235U in U at the start
of 1ife and is depleted to m].O%'235U in U at the end of its life.) In this
concept fission reactor spent fuel would be shipped from the reactor discharge
basin to a refabrication center. The spent fuel would be mechanically

VII-3
v=IIA

URANIUM MINE MILLING URANIUM CONVERSION

N
Q URANIUM ORE U304
N L) s '-..Egi l
TAILINGS,
ENRICHMENT
FUEL FABRICATION _ .. Fe Ry
FRESH FUEL ' ,
ENERGY
STORAGE
SPENT FUEL
RECYCLE
-‘lmmn C R RN OR
‘ DISPOSAL
REACTOR/GENERATING PLANT SURF

STORAGE OF SPENT
UNREPROCESSED FUEL

FIGURE VII-A-1. Uranium Nuclear Fuel Cycle

S-1IA

URANIUM
BLANKET

-

P ':\':

S BLANKET FABRICATION
URANIUM
i 5
STORAGE SPENT BLANKET HYBRID ENERGY
FOR

ULTIMATE DISPOSITION

FIGURE VII-A-2. Once-Through Hybrid Fuel Cycle

9=11IA

FISSION REACTOR

SPENT FUEL

S el

STORAGE OR DISPOSAL

CONVERSION FABRICATION URANIUM-FUEL

TG

THORIUM BLANKET REFRESH
URANIUM \ I CYCLE
NATURAL UF, @ - -; ;
% ENRICHED UF, \

ENRICHMENT

]

HYBRID

FIGURE VII-A-3. Refresh Hybrid Fuel Cycle

refabricated into fresh hybrid blanket module assemblies. This fuel would
then be re-enriched in the hybrid and, after an appropriate decay period,
returned to the fission reactor.

2. Reprocessing and Recycle of Fissile Materials

Denaturing a fissile isotope means diluting it with another isotope of

the same element to the extent that a nuclear weapon cannot be made from the
(235U and 233U) 238U

serves as the diluent. Until recently it was generally felt that plutonium

material. In the case of the fissile uranium isotopes

could not be denatured to render it unusable for weapons purposes. Recently,

238

it has been proposed by Allied-General Nuclear Services that Pu in suffi-

cient quantity in the plutonium can make the plutonium unusable for weapons

(2)

fixes to make the fuel cycle proliferation resistant include:

because of its high heat generation rate. Other technical and institutional

® Keeping the fissile and fertile materials together at all times (e.g.,
co-processed U-Pu) to dilute the fissile content to below
weapons-grade,

® Making the fuel highly radioactive (e.g., having highly radiocactive
materials in the fuel) to preclude handling,

(3)

Combining the above two in the CIVEX process, and

Restricting use to fuel cycle centers.

These concepts for the 233U and plutonium cycles are briefly described below.

a. Denatured 233U Cycle

235U

U supply can be alleviated through

This scenario assumes uranium and therefore its fissile component

is in short supply. The limitation of 232

233 (which is denatured) and thereby extend

utilization of thorium to generate
the supply of fissile material for fission reactors. The LWR thorium cycle
is shown in Figure VII-A-4, Raw materials bearing thorium are refined to
produce ThO2 which is mixed with enriched UO2 to fabricate ThO0,-U0, fuel for

2
233U which is

an LWR. The spent fuel is reprocessed to recover denatured
refabricated into new fuel. Since this is not a breeder cycle, an external
source of 233U is needed to sustain the system and allow it to grow. The

hybrid could be the external source.

VIii-7
8=-11A

ELECTRICAL

THORIUM ORE
J_,ENRICHED (THORITE) _
/ UF, SHIP RAW /

ENERGY ABRICATED ’
9 i EUEL MATERIAL /
ThOz ’-- " r
& A -
\ [ . ‘D- fl oy
MILLING AND THORIUM MINE
FUEL FABRICATION CONVERSION

& AND REFABRICATION I I
SPENT FUEL

VITRIFIED
HIGH-LEVEL

!

‘ RECOVER

fl:lSSILE MATERIAL

STORAGE

FISSILE MATERIAL

FIGURE VII-A-4.

RESIDUALS I

MONAZITE
SANDS
| Lo
EXTRACTION
STORAGE BYPRODUCT)

Thorijum LWR Fuel Cycle
The hybrid concept based on this scenario is shown in Figure VII-A-5,
Mined thorium is refined and a thorium blanket for the hybrid is fabricated.
Irradiating this blanket in the hybrids builds in 233
and denatured (238U added during reprocessing). The denatured uranium is
mixed with thorium during fabrication to produce LWR fuel. Once the spent
fuel is discharged from the LWR, the steps shown in Figure VII-A-5 would be

U which is reprocessed

followed.

The isotope 232U builds up in these cycles to the point where the radia-

tion levels are sufficient to require massive shielding during handling and
processing.(4) The requirements for shielding are perceived as adding pro-
liferation resistance to the fuel cycle.

b. Plutonium Recovery and Recycle

As mentioned above it has been suggested that sufficient addition of

Pu in the plutonium can render it unusable for weapons purposes. The
238 237Np from the

238

level of Pu can be increased by recovering uranium and
spent fuel. The isotope 236U builds up during irradiation of the fresh UO2

fuel in the LWR. Subsequent irradiation of the 25U and 2°'Np produce
238 (4)

a hybrid can contain significant fractions of

that the plutonium produced in
236 238Pu

Pu in the plutonium. It has been: shown
Pu and

Since LWRs do not convert a sufficient amount of plutonium to fuel them-
selves, an external source of plutonium is needed to sustain the system and
allow it to grow. As shown in Figure VII-A-6 the hybrid could be the external
source of proliferation resistant plutonium. The sources of uranium include:
mixed natural, depleted uranium from the enrichment plants and/or the uranium
recovered in reprocessing spent U02 LHR fuel.

The hybrid fuel cycle in this scenario is similar to the fission breeder
cycles in that fission reactor fuel supply requirements are extended by con-
verting 238U to fissile plutonium. The plutonium produced in hybrid blankets
could be subject to the same restrictions as that produced in fission reactor
fuels, namely, it would be rendered proliferation resistant at the same stage

as fission reactor fuel.

VII-9
OL-TIA

THORIUM BLANKET
FABRICATION
N |
|
LWR
SPENT FUEL
\ TO
‘ REPROCESSING
L_] AND RECYCLE
l [ 4.....
ENERGY
BRED
BLANKET

=

THORIUM
BLANKET

l HYBRID

E=g s

[

DENATURED REFABRICATION DENATURED REPROCESSING STORAGE
UQ; (ThO,) URANIUM FOR
ULTIMATE
DISPOSITION

FIGURE VII-A-5. Thorium Hybrid Fuel

Cycle
LL-TIA

BLANKET URANIUM
FABRICATION BLANKET HYBRID

o7l

URANIUM

UO; - PUOz
RECYCLE FUEL
BRED
BLANKET
COPROCESSED
PLUTONIUM-URANIUM fl
]
4y /= I__LI
[ [r
REFABRICATION REPROCESSING
HIGH-LEVEL WASTE FOR
ULTIMATE
DISPOSITION

FIGURE VII-A-6. Plutonium Recycle

Table VII-A-1 shows a list of specific driver/blanket fuel cycie combina-
tions. Each of these fuel cycles will be characterized and placed into a
non-proliferation perspective relative to a once-through or throwaway fuel
cycle. Included in this characterization is a discussion of fuel manage-
ment strategies and overall fuel requirements. Fuel cycle facility des-
criptions can be made once these fuel management strategies and fuel mass
flows have been established.

TABLE VII-A-1. Driver/Blanket Fuel Cycie Combinations

Driver Once-Through Pu-Recycle Refresh Pu Catalyst
Tokamak ° ° ° °
Mirror ° © ° °
Theta-Pinch ° - ° ° °
Laser ° ° ° °

B. FUEL MANAGEMENT STRATEGIES

The blanket module management scheme is to place fertile fuel into the
hybrid reactor and expose the fuel to a specified burnup. After the maximum
burnup level is achieved the spent module is removed and placed into a hot
cell operations area. Here the fuel is unloaded and prepared for shipment to
a reprocessing plant where fissile fuel created in the hybrid is separated
from the spent fuel. The thermal output of the blanket is increased as it
resides in the hybrid due to energy multiplication resulting from the fission-
ing of fissile material created in the hybrid blanket. The hybrid reactor
should be operated such that the blanket management tends to minimize the power
variations within the blanket modules. |

1. Tokamak Hybrid Reactor

The Tokamak Hybrid Reactor blanket modules are divided into 4 fuel
management regions. The tokamak blanket consists of 60 slices with each
slice made up of 11 modules (660 total modules). The blanket 1ife is taken
to be 4 years so that each year 165 modules or 1/4 of the blanket fuel is

VII-12
removed and replaced with fresh fuel. Table VII-B-1 gives some nertinent fuel
management information.

TABLE VII-B-1. Tokamak Hybrid Fuel Management Data

Number of Fuel Management Regions 4
Blanket Lifetime 4 years
Plant Capacity Factor 0.75

- Maximum Exposure 9.6 MWy/m?
Blanket Replacement Time 35 days
Fuel Management Interval 1.3 years

The capacity factor for the Tokamak Hybrid Reactor is influenced by several
factors, There are many subsystems that require maintenance. Table VII-B-2
lists some of these systems.

TABLE VII-B-2. Reactor Subsystems

Coolant Loops

Cryopump, N, He supplies
Divertor plates

Neutral Beam injectors

o hRhw NN -

Inspection of toroidal field coils, poloidal field

coils, fluid Tines and manifolds.

6. Tritium cleanup systems, coolant purification, and
sieve getter beds

7. Shielding

For the purposes of this study it is assumed that maintenance of these
systems is accomplished during blanket module replacement. Unscheduled
maintenance results from first wall burn-through, coolant leaks, isolation of
failed modules or any abnormal operating conditions, The unscheduled

VII-13
maintenance is taken to be approximately 50 to 60 days/year, A portion of
this time is for some scheduled maintenance not performed during blanket
change operations. The blanket replacement outage will last about 35 days

(1.5 days/slice plus 5 days for shutdown, decay heat removal, and tritium
outgassing).

Shown in Table VII-B-3 are the initial fertile fuel, structure, tritium
breeder, and reflector material requirements.

The blanket configuration for the Once-Through and Pu-Recycle fuel
cycles both use uranium carbide as the fertile material, while the Pu-Catalyst
fuel cycle has a converter region of uranium dioxide and plutonium dioxide
and a fertile region of thorium carbide. The refresh fuel cycle makes
use of uranium dioxide. The isotopic feed for each of these fuel cycles

is natural uranium (0.72% 235U).

In the Pu-Catalyst fuel cycle the converter
region uses a mixed-oxide (UOZ/PuOZ) fuel matrix with an equilibrium
concentration of ~ 8% Pu02. The tritium breeding material, L120, is enriched
to 90% 6L1. Both the fertile fuel and tritium breeder are contained in
stainless steel rods. A graphite reflector region is also incorporated into

the blanket module.

Using the fuel management information presented in Table VII-B-3, the
annual charge and discharge fuel quantities can be determined. The 30-
year requirements are given in Tables VII-B-4 - VII-B-6. This isotopic
information was obtained through the use of the ORIGEN computer code.(s)
ORIGEN is a point depletion code which solves equations of radioactive
growth and decay. The ORIGEN program considers (n,y),{(n,2n), (n,p) and
(n,a) reactions for 1light elements and structural material. The actinides
have (n,y), (n, fission), (n,2n) and (n,3n) reactions included. The total
fuel mass flows for each of the four cycles are shown in Figures VII-B-1 to
VII-B-4. Listed below are some of the assumptions used in developing
these fuel flows.

a. Once-Through Assumptions

(a) Li Content in crude ore - 5% Li/ore

VII-14
GL-TIA

TABLE VII-B-3. Tokamak Hybrid Reactor Initial Material Requirements

Tritium

Reactor/Fuel Cycle Fuel Structure Breeding Material
Tokamak/Once-Through  317.8 MTUC 170.9 MTss(@) 123.62 MT LiZO(b)
Tokamak/Pu Recycle
to Thermal Reactors 317.8 MTUC 170.9 MTSS 123.62 MT LiZO
Tokamak/Refresh 234.0 MT U0, 170.9 MTSS 123.62 MT Li,0
Tokamak/Pu-Catalyst 113 MT UO2

223 MT ThC 170.9 MTSS 85.8 MT L120

9 MT PuO2

(a) Stainless steel amount includes cladding and module structure.
(b) Tritium breeding material, Li»0, enriched to 90% 6L1.
(c) Graphite is for reflector region of module. ’

Other

55.97 Mrc(C)

55.97 MTC
55.97 MIC

55.97 MTC
9L-TIA

Year

—r
QWO NOOUVIBWN-—-

— ) —t —d
W N —

15

TABLE VII-B-4.

FUEL MANAGEMENT CHARACTERISTICS

Reactor Type Tokamak Hybrid

Once-Through and Pu-Recycle Fuel Charge Data

Natural Uranium Carbide

Fuel Type
Reactor Fuel Charge Data
Annual Heavy Metal, Thorium, Uranium, kg Plutonium, kg
Capacity Factor, % kg kg 233 234 235 236 238 239 280 281 282

75 302,540 0 0 0 2178 0 300,345 0 0 0 0
75 75,635 . . . 544 - 75,086 . .« . .
75 75,635 544 75,086

75 75,635 544 75,086

75 75,635 0 0 0 544 0 75,086 0 0 0 0

[1-TIA

-
(1]
-]
1

WONMTN DBWN -~

TABLE VII-B-5.

FUEL MANAGEMENT CHARACTERISTICS

Reactor Type  Tokamak Hybrid

Once-Through and Pu-Recycle Fuel Discharge Data

Fuel Type Natural Uranium Carbide

Reactor Fuel Discharge Data

Heavy Metal, Thorium, Uranium, kg
kg kg 233 234 235 236 238
74838 0 .006 .072 487 27.2 73593
74843 . .014 .18 452 58 72880
74840 . .026 .38 419 86 72168
74829 . .036 .62 387 114 71455
74625 0 036 .67 387 114 71455

Plutonium, k

233

1887

20 241
4.8 .044
8.7 .14
3.4 .30
7.5 .46
17.5 .46

24z

Fission

Products, kg

190.8
381.7
572.6
763.5

Other Isotopics, kg

Pa-233  Np-237 Am-241 Cm-242
0 55 ~0 0
. 107 .006
. 156 .029
. 204 .088  .00136
208 .088 0
8L-TIA

TABLE VII-B-6. Pu-Catalyst Fuel Charge Data

FUEL MANAGEMENT CHARACTERISTICS

Reactor Type Tokamak Hybrid Reactor

Fuel Type U02/Pu02 Convertor Reaion

ThC Breedina Region
Reactor Fuel Charge Data
Annual Heavy Metal, Thorium, Uranium, kg Plutonium, kg'?)

Year Capacity Factor, % kg kg 233 234 235 236 238 239 240 241 242

75 325,674 212,033 0 0 717.2 0 98,889 7968 3498 1870 699
75 81,418 53,008 . -« 179.3 » 24,722 1992 874 467 174
75 81,418 53,008 . - 179.3 » 24,722 1992 874 467 174
74 81,418 53,008 . - 179.3 - 24,722 1992 874 467 174

O ONOOUTHE W —

30 75 81,418 53,008 0 0 179.3 0 24,722 1992 874 467 174

(a) LWR Discharge Plutonium
LITHIUM CRUDE ORES WATER LRANIUM ORES

fiAg%L,i“(;RES l l L l 40,212 MT CRUDE ORES

Li MINING DEUTERIUM U MINING
AND MILLING PROCESS ING AND MILLING
. D: 25.7 kalyr
695.7 MT Li v et 76.4 MT U
Li ENR ICHMENT BFEEI‘IIILLEET
FABRICATION

FABRICATION
123.6 MT Lix0
(57.4 MT Li)
75,6 MT U
I

_ TOKAMAK
Li;0-30.9 MT HYBRID REACTOR

> TOTAL SPENT FUEL
38 kg Tlyr U: 72 MT
RECYCLE T " ACTIVATED SS: 9.7 MT
BRED T 165 SPENT | Pu: 1.95> MT .
F b 7635 kqivr ACTINIDES: 887 kg
= 45 kglyr MODULES | F-F.: 703.0Kglyr
SEPARATION OF | |

| RADIOACTIVE
D, Li, T, He WASTE DI1SPOSAL

T.B.R. = L 19
6.8 kglyr TO
STORAGE
TRITIUM FOWER UNBURNED FUEL
STORAGE 1000 Mwe D, T, Li
D: 231kqglyr
T: 348 kglyr

FIGURE VII-B-1. Tokamak Hybrid Reactor Fuel Flow -
Once-Through Fuel Cycle

VII-19
LITHIUM CRUDE ORES WATER URANIUM ORES
14, 646 MT 40,212 MT CRUDE U ORE
NAT. Li ORES l l £ &
Li MINING DEUTERI UM U MINING
AND MILLING PROCESSING AND MILLING
695.7 MT Li D: 25.7 kglyr 76.4 MT U
T: 38.2 kglyr
A
Li ENRICHMENT FERTILE
FABRICATION BLANKET
FABRICATION
123.6 MT Li0
7.4MTLD 75,6 MT L “RECYCLE U?
_ 3 165 SPENT
Li20-30.9 MT TOK AMAK MODULES MOL ULE
HYBRID REACTOR REPROCESSING
~38 kg Tiyr '
RECYCLE U: 72 MT
TRITIUM Pu: 1.95MT BRED FUEL
F.P.: 7063.5kglyr P 1950 kg
'?F;EDKT, ) ACTINIDES. 57 kg , (Kp. oy
- P RGY ACTIVATED SS: 9.7 MT
SEPARATION OF & RADIOACTIVE
D, Li, T, He WASTE DI SPOSAL
TB.R.-1.19
6.8 kq/yr TO
STORAGE I v {
TRITIUM POWER UNBURNED FUtL FISSILE FUEL
STORAGE 1000 MW, u T, STORAGE
B 23i kgt
T 348 ky'yr 3 g PuUlR
RECYCLE OPTION 514 ky ¢3"Pu R
LIGHT WATER LWR
REPROCESSING 14— REACTOR RL-FABRICATION
SPENT FUEL . /
TOTAL FUEL
Pu. 1,58 MT
F.P. AND 3RX}MLH
RADIOACTIVE (NAT.
A DES: 1
WASTE DISPOSAL [ CTINIDES: 113 MT
U= 0.303MT POWLR
Pu= 0.0158 MT 1300 o
F.P. AND

ACTINIDES = 1.13 MT

FIGURE VII-B-2.

VII-20

Tokamak Hybrid Reactor Fuel Flow - Pu-Recycle
LITHI UM CRUDE ORES WATER THOR{ UM URANIUM ORES
10, 155 MT 13237 MT U30g ORES
NAT. Li ORES L l 1126 MT ThGy ORES
Li MINING DEUTERIUM U, Th MINING
AND MILLING PROCESSING AND MILLING
. D: 25.7 knjiyr 53.5 MT Th
482.4 MT Li T. 3.2 kayr 25.1MT U
EXTERNAL Pu:
9 MT
. FERTILE
Li ENRICHMENT BLANKET ‘T
FABRICATION FABRICATION
: 2.25MT Pu
85.8 MT Liz0 24.9 MT U
(39.8 MT Li) 53 0 MT T RECYCLE
i U, Th, Pu
Yy Vv l 165 SPENT
Li20-21.45 MT TOK AMAK MGDULES MODULE -
HYBRID REACTOR ' REPROCESSING {—
~ 38 kgfyr
RECYCLE
TRITIUM
SEPARATION OF RADIOACTIVE BREu FUEL
D, Li, T, He |45 kglyr WASTE DISPOSAL U-233:
TRITIUM 6.8 kglyr TO
B.R.-1.18 STORAGE [__ & rJ
TRITIUM POWER UNBURNED FUEL FISSILE FUEL
STORAGE 1835 MW, D, T, Li STORAGE
D: 231 kgfyr
T 348 kylyr 212.7 kg U-233/PWR
RECYCLE OPTION
—b
| 458 kg U-233
LIGHT WATER LWR
REPROCESSING REACTOR RE-FABRICATION
U: 7.4 MT 30.5 MTH.M. 4
: MAKEUP FUFL
Th: 22 MT SPENT Rl
Pu: 105 kg FUEL Th: 22.1 MT
F.P.+ ACTINIDES U: 7.9MT
RADIOACTIVE 1053 kg
WASTE D} SPOSAL
F.P.+ ACTINIDES - 1053 kg Lo
Th: 0,22 MT e

U: 0.074 M1
Pu: 0.001 MT

FIGURE VII-B-3,

Tokamak Hybrid Reactor Fuel Flow -

Pu-Catalyst Fuel Cycle

VII-2]
LITHIUM CRUDE ORES

WATER URANIUM ORES
14,646 MT
NAT. L1ORES l l J l 27415 MT U30g ORES
Li MINING DEUTERIUM L MINING
AND MILLING PROCESS ING AND MILLING
695.7 MT Li D: 25.7 kglyr
T. 38.2 kglyr 52.1MT U
Li ENRICHMENT FERTILE
FABRICATION BLANKET
FABRICATION
l 123.6 MT Liz0
(57.4 MT Li) SL6 MT
Iy
Lix0 - 30,9 MT TOKAMAK
HYBRID REACTOR | 823 MiWe)
~38 kg Tlyr | RECYCLE T‘
TRITIUM
44.7 kglyr
SEPARATION OF |
0. G T, He ——t - — — — — — RE-FABRICATION
T B.R. = 117 Pu = 1387 kglyr
6.5 kgfyr F.P. = 779 kglyr DEPLETED FUEL
TO STORAGE TOTAL FUEL: 51 MT
"ENRICHED FUEL"
TRITIUM AND
UNBURNED FUEL
STORAGE
D: 231kglyr
T: 348 kqlyr ¢
LIGHT WATER » RADIOACTIVE
REACTOR WASTE DISPOSAL

FIGURE VII-B-4.

Tokamak Hybrid Reactor Fuel Flow -
Refresh Fuel Cycle

VII-22
(b) D-T requirements represent quantity burned.

Hybrid will become self-sufficient after a certain time

(B.R. = 1.19) - external tritium supply unnecessary

(¢) Uranium assay - 0.2% U/crude ore (U308)

(d) 75% Plant factor used.

(e) 4 year fuel cycle - 1/4 modules reolaced each year.

(f) D-T processing losses not considered in unburned fuel quantity.

(g) Lithium is 90% ®Li enriched. The amount of natural Li required

to obtain this enrichment is 12 x amount desired.

(h) It is assumed that Li,0 breeding pins are re-used. A continuous

supply is unnecessary since the quantity of 6L1' consumed,

76 kg/year, is small compared to the initial inventory (124 MT
L1,0, 57 MT Li). The °
consumed per fusion neutron.

Li consumption is based on one atom

(i) The module stainless steel structure is re-used. The only
components replaced are the fertile fuel pins. Occasionally a
tritium breeding rod may be replaced. The graphite reflector
region is also recycled each year,

(j) The tritium extraction is a batch method whereby the pins are
heated and the outgassed tritium collected. This process will
occur in a separate vacuum containment and hot cell module
maintenance area.

(k) The fabrication process losses are taken to be 1% of the feed
inventory. The milling process is assumed to recover 95% of
the available U or Li in the ores.

b. Pu-Recycle to Thermal Reactor

This fuel cycle is based upon the same assumptions that made up the
Once-Through fuel cycle plus the followign assumptions:

(a) The reprocessing losses are taken to be 1% of initial feed
material (spent fuel).

(b) Although separate reprocessing facilities for the LWR and Hybrid
are shown in the diagram, these processes would probably be
carried out at a common reprocessing facility. The Tokamak Hybrid

VII-23
c.

d.

(c)

fuel rods are similar to the LWR fuel rods. The fabrication oflv
L120 and UC fuel rods could also be achieved at the same facility.
In the fuel cycle the uranium separated from the bred plutonium

is recycled back to be used as fuel material in fresh fertile
module elements.

The LWR data (Pu requirements, power, etc,) is based upon the
Combustion Engineering PWR Pu/U fuel cycle. Averaged over 30 years
the charge to the reactor is 854 kg and discharge is 519 kg.(s)
This leaves ~334 kg makeup which is supplied by the hybrid

reactor.

Pu Catalyst Fuel Cycle

An initial amount of PuOj is required for start-up (8% by weight

in the converter region).

U and Th are recycled back to the hybrid modules.

LWR fuel cycle used is the PWR cycle: denatured (20%) U-235/Th
fuel with U-233 self-generated recycle and U-233 makeup from the
hybrid reactor. This is a Combustion Engineering PWR fuel cyc]e.6
The thorium ore assay is 5% ThO /monazite sands.

Refresh Fuel Cycle

(a)
(b)

A direct exchange of fuel between Hybrid and PWR.
Replacement schedule based upon number of years required
to obtain Pu/U enrichment of ~2.7%.

Re-fabrication is mechanical separation of fuels and cladding/
assemblies etc.

PWR based upon Combustion Engineering Pu/U reactor fuel cyc]e.(s)
The hybrid produced mixed-oxide fuel has Pu-239 quality ~ 90-98%
as compared to normal Pu-239 feed quality of ~ 55%.

The Pu-recycle fuel cycle utilizes the Tokamak Hybrid both as a power
producer and fissile fuel producer, The fissile output, 1950 kg Pu-239/year,
is used in a Pu/U fueled PWR. The reference PWR used in this study is a

Combustion Engineering design producing 1300 MWe. The 30-year average annual
Pu makeup requirement is 334 kg/year assuming recycle Pu from PWR spent fuel

VII-24
reprocessing is used. The Tokamak Hybrid could theoretically support the
annual fissile makeup requirements for 5-6 PWRs.

The Pu-catalyst fuel cycle, also operating in the reprocessing mode,
consists of the Tokamak Hybrid Reactor coupled with a PWR operating on a
denatured U-233 fuel cycle. This reactor is also a Combusion Engineering
fuel cycle design producing 1300 MWHe. Based on the annual U-233 makeup
requirements and assuming the PWR has self-generated U-233 recycle the hybrid

breeder could support 13 to 14 PWRs.

The refresh fuel cycle uses the annual discharge fuel from the Tokamak
Hybrid to fuel a PWR. 1In this scenario the hybrid spent fuel 1is mechanically
separated and refabricated into PWR fuel elements and assemblies. No
chemical separation or reprocessing is required. An initial feed of natural.
uranium exposed to a four-year burn cycle would result in a fissile Pu/U
enrichment of 2 to 3%. It is possible that the annual discharge of fuel from
the Tokamak Hybrid could support two PWRs in this manner.

2. Mirror Hybrid Reactor

The Mirror Hybrid Reactor is a collection of 16 "peel" shaped segments
located around a spherical plasma chamber. Each of these slices consists of
approximately 45 modules depending on the location of the segment (near a
beam port or fan hole). There are 600 modules located around the plasma

‘cavity. The Mirror blanket has been divided up into 4 management regions.
Table VII-B-7 lists the important fuel management information.

TABLE VII-B-7. Mirror Hybrid Fuel Management Data

Number of Fuel Management Regions 4

Blanket Lifetime 4 years
Plant Capacity Factor 0.75
Blanket Replacement Time 35 days
Fuel Management Interval 1.33 years
Maximum Blanket Exposure 6.7 Mwy/m2
Number of Blanket Segments ' 16

Number of Blanket Modules 600

Number of Blanket Modules Replaced Each Outage 150 modules/year

VII-25
The maintenance of the reactor subsystems will occur mainly during blanket
replacement outages but some maintenance or repair will occur that is
unscheduled as was the situation in the Tokamak Hybrid Reactor. The Mirror
Hybrid Reactor segments are quite massive, ~ 40 MT total weight, and

although the crane hoisting system presents no unusual problems, it will

take Tonger to gain access to the modules because of a massive concrete plug.
One-fourth or 4 segments will be replaced during blanket change outage. It
will take 5 to 8 days to remove a segment, disassemble the module, and replenish
them with fresh fuel. A total blanket change time of 35 days, same as the
tokamak, was assumed for the Mirror Hybrid Reactor. The total unscheduled
outage time was taken to be 50 to 60 days/year.

Table VII-B-8 presents the initial loading of fertile fuel, structure,
tritium breeder, and graphite reflector. The fuel and breeder material
choices for the Mirror Hybrid Reactor are the same as those of the
Tokamak Hybrid.

The 30-year plant lifetime requirements are given in Tables VII-B-9 to
VII-B-11. The fuel mass flow balance is based upon these quantities and
their isotopic content. Figure VII-B-5 shows the fuel mass balance for the
Pu-Catalyst fuel cycle. This fuel cycle is capable of supplying the fissile
makeup quantities for 5 to 6 PWRs. It should be noted that because of the
scope of the NASAP study the driver blanket combinations were not optimized
except for the Tokamak Hybrid Reactor. The Once-Through and Pu-Recycle blankets
exhibit low power outputs and thus are not viable fuel cycles. The Mirror
Refresh fuel cycle requires a 10-year burn cycle for the Pu/U enrichment to
be ~3%.

a. Laser Hybrid Reactor

The Laser Hybrid Reactor has a cylindrical plasma chamber with fertile
and tritium breeding regions surrounding this cavity. There are also
top and bottom tritijum breeding blankets. The radial fertile blanket has
rows of fuel elements and is divided into eight segments. Table VII-B-12
gives some important fuel management information.

VII-26
LZ-1IA

TABLE VII-B-8. Mirror Hybrid Reactor Initial Material Requirements

Tritium
Reactor/Fuel Cycle Fuel Structure Breeding Material Other

#irror/Once-Through 303 MT UC 162 #7 s5(3) 116.5 MT Ligo(b) 53 M7 c(¢)
Mirror/Pu Recycle to
Thermal Reactors 303 MT UC 162 MT SS 116.5 MT Li,0 53 MT C
Mirror/Refresh 223.5 MT U0, 162 MT SS 116.5 MT Li,0 53 MT C
Mirror/Pu Catalyst 108 MT UO2

212 MT ThC 162 MTSS 81 MT Li,0 53 MT C

8.7 MT PuO,

(a) Stainless steel amount includes cladding and module structure.
(b) Tritium breeding material, Li20, enriched to 90% 6Li.
(c) Graphite is for reflector region of module.
TABLE VII-B-9. Laser Hybrid Fuel Management Data

Number of Fuel Management Regions 4

Blanket Lifetime 4 years
Plant Capacity Factor 0.75
Blanket Replacement time 35 days
Fuel Management Interval 1.33 years
Maximum Blanket Exposure v 6 MAy/m2
Number of Blanket Segments 8

Number of Blanket Segments Replaced 2

The Laser Hybrid Reactor fuel requirements will be based on a four-year
burnup cycle. The unscheduled and scheduled maintenance outage times, 90 days/
year, results in a plant capacity factor of 0.75. Table VII-B-13 gives the
initial loading of fertile fuel, structure, tritium breeder and graphite
reflector.

The 30-year plant lifetime fuel requirements are given in Tables VII-B-14
to VII-B-16. The fuel mass flows for the blanket management schedme discussed
previously is given in Figures VII-B-6 to VII-B-8. The Once-Through, Pu-Recycle
and Pu-Catalyst fuel cycles are shown in these fiqures. The annual discharge of
Pu in the Pu-Recycle case, 1323 kg Pu-239, can be remotely refabricated into

mixed-oxide PWR fuel elements. This plutonium could support 3 to 4 PWRs each
year. The Pu-Catalyst fuel cycle produces 2584 kg U-233 each year. The
fissile output is capable of supporting 9 to 10 PWRs. The refresh fuel cycle
was not optimized with the result that a 10-year burn cycle is required to get
a v 3% Pu/U enrichment.

b. Theta-Pinch Hybrid Reactor

The Theta-Pinch Hybrid reactor is made up of 200 2.5 meter long cylindrical
modules. The fuel and tritium breeder elements are arranged parallel to the
plasma cavity axis. The fuel management information is listed in Table VII-B-17.

VII-28
62 -11IA

ce=11A

TABLE VII-B~-10. Pu-Recycle Fuel Charge Data

FUEL MANAGEMENT CHARACTERISTICS

Reactor Type  Mirror Kybrid Reactor -

Fuel Type Natural Ursnium Carbide

Reactor Fuel Charge Data

Annual Heavy Metal, Thorium, Uranium, kg Plutonium, kg
Year Capacity Factor, % kg kg 233 234 235 236 238 239 280 241 242

75 288,800
75 72,200
75 72,200
75 72,200

0 207¢ 0 286,705
. 520 71,676
. 520 71,676
. 520 71,676

[ S T T B I - ]
«a s s a &

«a s & s s & O

s o s » 2 s O

0

— OW RN W N -

-

30 75 72,000 0 0 0 520 0 71,67 0 0

-
)
o
B

WO~NOREWMN—

0E-IIA

TABLE VIT-B-11, Pu-Recycle Fuel Discharge Data

FUEL MANAGEMENT CHARACTERISTICS

Reactor Type Mirror Hybrid Reactor

Fuel Type Natural Uranium Carbide
Reactor Fuel Discharge Data
Heavy Metal, Thorium, Uranium, kg Plutonium, kg Fission Other Isotopics, kg

kg kg 233 234 235 236 238 239 240 241 24z Products, kg Pa-233 Np-237 Am-241 (m-242
71,519 0 .006 .07 468 26.1 70,706 199 2 .02 0 66.4 0 52 ) 0
71,146 . .013 .17 434 55.5 70,022 395 5.6 .08 . 132.7 . 101 .006 .
70,769 . .025 .36 402 83.2 69,337 589 11.2 .20 . 199.1 . 148 .027 .
70,392 . .034 .59 372 109 68,652 781 18.4 39 - 265.5 . 194 .084 .0013

- . . . . -

- - - - . . - . . - . * - - - -

70,392 0 .034 .59 372 109 68,652 781 18.4 .39 0 265.5 0 194 .084 .0013

LITHI UM CRUDE ORES WATER THORIUM, URANIUM ORES
9601 MT 12631 MT U30g ORES
NAT.LiORES & & J l 1074 MT ThO; ORES
Li MINING DEUTERI UM U, Th MINING
AND MILLING PROCESSING ANE MILLING
D: 9.1kglyr 20 MT U
456 MT Li T. 13.0kglyr 51 MT Th
EXTERNAL Pu
8.7 MT
Li ENRICHMENT FERTILE |
BLANKET
FABRICATION FABRICATION
U: 23.8 MT
81 MT Liz0 Th: 50.4 MT RECYCLE
(37 MT Li) P 2.17MT U
» Th
1 i Pu0,
SPENT
Li0 - 20.2MT MIRROR HYBRID |t QRULES MODULE  |—
| ol REACTOR REPROCESSING F—
~13 kglyr
TRITIUM
RECYCLE
SEPARATION OF RADIOACTIVE BRED FUEL
DT, Li, He | lokgiyr WASTE DI SPOSAL 1575 kg U-233/yr
BRED T
TRITHUM 2.4 kglyr STORAGE
B.R.-1.18 [—— l —
TRITIUM POWER UNBURNED FUEL FISSILE FUEL
STORAGE 544 MW, D. T, Li STORAGE
D. 1000 kgyr
T. 1497 kglyr )
RECYCLE. 458 kq U-233 212 kg U-233PaR
Y
LIGHT WATER | LWR
REPROCESSING REACTOR RE-FABRICATION
U 7.4MT , y
Pu: 105 kg TOT@EZ%aT‘OAD: MAKEUP FUEL
F_P.+ ACTINIDES: 1053 ky : —
RADIOACTIVE Th: 22 MT
WASTE DISPOSAL [ .
F.P.+ ACTINIDE - 1053 kg 12&“&%
Th-0.2MT  Pu-0.001 MT ¢

U=0.074 MT

FIGURE VII-B-5.

VII-31

Mirror Hybrid Reactor - Pu-Catalyst
¢e-TIA

Year

LLoO~NOUTLE WA —

10

TABLE VII-B-12.

FUEL MANAGEMENT CHARACTERISTICS

Reactor Type Mirror Hybrid Reactor

U02/Pu02 Convertor Region

Thorium, Uranium, kg Plutonium, kg

Pu-Catalyst Fuel Charge Data

(a)

Fuel Type
ThC Breeding Reaion
Reactor Fuel Charge Data
Annual Heavy Metal,
Capacity Factor, % kg kg 233 234 23b
75 310,370 201,570 0 0 685
75 77,592 50,392 . . 171
75 77,592 50,392 . . 171
75 77,592 50,392 . . 171
75 77,592 50,392 0 0 171

(a) LWR Discharge Plutonjum

2% 238 B 200 Al @

0 94.509 7588 3332 1881 666
+ 23,627 1897 B33 445 166
23,627 1897 833 445 166
23,627 1897 833 445 166

0 23,627 1897 B33 445 166
EE-TIA

TABLE VII-B-13. Laser Hybrid Reactor Initial Material Requirements

Reactor/Fuel Cycle Fuel Structure Breeding Material Other
Laser/Once-Through 488 MT UC 249 MT SS(a) 169 MT L120(b) 75 MT C(C)
Laser/Pu Recycle to
Thermal Reactors 488 MT UC 249 MT SS 169 MT L120 75 MT C
Laser/Refresh 360 MT UO2 249 MT SS 169 MT L120 75 MT C
Laser/Pu Catalyst 177.5 MT UO2

328.5 MT ThC 249 MT SS 103 MT L120 75 MT C
14 MT PuO2

(a) Stainless steel amount includes cladding and module structure.
(b) Tritium breeding material, Li»0, enriched to 90% 6Li.
(c) Graphite is for reflector region of module.
TABLE VII-B-14. Once-Through and Pu-Recycle Fuel Charge Data

FUEL MANAGEMENT CHARACTERISTICS

Reactor Type Laser Hybrid Reactor

Fuel Type Natural Uranium Carbide

Reactor Fuel Charge Data

vE-TIA

—<
)
-1
=

—
— OOV ONOUIEWN

—t ) ot ad el ) mnd e
OO~NOOdWN

Annual Heavy Metal, Thorium, Uranium, k Plutonium, kg
kg Z3 7z 25 2% 2B 239 20 2 242

Capacity Factor, % kg 234
75 464,576 0 0 0 3345 0 461,206 0 0 0 0
75 116,144 . . 836 115,301 . . . .
75 116,144 . . 836 + 115,301 . . . .
75 116,144 . . 836 - 115,301 . . . .
75 116,144 0 0 0 836 0 115,301 0 0 0 0

GE-TIA

TABLE VIT-B-15. Once-Through and Pu-Recycle Fuel Discharge Data

FUEL MANAGEMENT CHARACTERISTICS

Reactor Type Laser Hybrid Reactor

Fuel Type Natural Uranium Carbide

Reactor Fuel Discharge Data
Heavy Metal, Thorium Uranium, kg Plutonium, kg Fission Qther Isotopics, kg
kg kg 233 234 235 236 238 239 240 241 242 Products, kg Pa-233 Np-237 Am-241 (Cm-242

<
)
o
=

115,000 0 .02 .28 752 42 113,653 326 3.27 .03 0 129.5 20 85 ~0 0
114,419 . .04 .54 698 89 112,552 648 9.22 .13 . 259 . 164 .045

113,844 . .056 .96 646 133 111,451 966 18.3 .33 . 388 . 241 .13 .
113,268 . .069 1.4 597 176 110,351 1280 30.2 .64 . 518 . 315 .31 .002

- . - . -

OOV WA =

30 113,268 0 069 1.4 597 175 110,351 1280  30.2 .64 0 518 0 315 .31 .002
9e-TIA

TABLE VII-B-16.

FUEL MANAGEMENT CHARACTERISTICS

Reactor Type Laser Hybrid Reactor

Fuel Type UOZ/PUO2 Convertor Region

ThC Breeding Reaion

Annual Heavy Metal,

Year Capacity Factor, % kg

Reactor Fuel Charge Data

OWONIN B WN—

75 491,240
75 122,810
75 122,810
75 122,810

75. 122,810

(a) LWR Discharge Plutonium

Pu-Catalyst Fuel Charge Data

Thorium, Uranium, kg Plutonium, kg(a

kg 33 238 235
312,340 0 0 1126
78,085 . . 281.5
78,085 . 281.5
78,085 . .
78,085 0 0 281.5

236 238 239 280 241 242

0 155,334 12,515 5495 2938 1099
- 38,833 3,128 1373 734 274
- 38,833 3,128 1373 734 274

0 38,833 3,128 1373 734 274
URANIUM ORES
L 61746 MT U30g ORES

U MINING
AND MILLING

117.3 MTU

FERTILE
BLANKET
FABRICATION

LITHIUM CRUDE ORES WATER
20031 MT
NAT. Li ORE |
Li MINING DEUTERIUM
AND MILLING PROCESSING
951 MT Li D: 19.8 kglyr
T. 29.7 kglyr
Li ENRICHMENT
FABRICATION
169 MT Li,0
(78.5 MT Li) 116 MTU
v vy
Li50 - 42.2 MT LASER
HYBRID REACTOR
RECYCLE
TRITIUM

SEPARATION OF [€——————ri

TOTAL SPENT FUEL

U: 111 MT ACTIVATED SS.: 14.3 MT
Pu: 1.3 MT ACTINIDES: 327 kg
F.P.: 513 kq/yr

| »| RADIOACTIVE

D, T,Li, He 35.3 kglyr WASTE DISPOSAL
BRED T
TRITIUM
B.R.-1.19 5.6 kg Tlyr
w STORAGE v
TRITIUM POWER UNBURNED FUEL
STORAGE 940 MW, D,T,Li
D: 376 kglyr
T: 564 kglyr

FIGURE VII-B-6. Laser Hybrid Reactor - Once-Through Fuel Cycle

VIiI-37
LITHIUM CRUDE ORES WATER URANIUM ORES

20031 MT 61746 MT U30g ORES
NAT. Li ORES l L l l
Li MINING DEUTERIUM U MINING
AND MILLING PROCESSING AND MILLING
D: 19.8kylyr
951 MT L T. 29.7kghyr 117.3 MTU
FERTILE
Li ENRICHMENT BLANKET 4
FABRICATION FABRICATION
169 MT Lip0
(78.5 MT Li) 116 MTU RECYCLE U
l ‘ SPENT
. LASER MODULES MODULE —
Li20 -42.2 MT HYBRID REACTOR REPROCESSING
SEmma—
TRITHUM gu “IIBM&T
YC L B
RECYCLE e 518 kalyr RED FUEL
ACTINIDES: 327 kg
" ACT IVATED
SEPARATION OF [0, SS: 143 MI RADICACTIVE 1323 kg Pu-239/yr
D, T, Li, He = Kg7y WASTE DISPOSAL
- BRED T
o | oae
B.R.-1.19
) 4 v
TRITIUM POWER UNBURNED FUEL FISSILE FUEL
STORAGE 940 MW, D, T Li STORAGE
D: 376 kalyr
RECYCLE: 514 kg Pu-239
LIGHT WATER LWR
REPROCESSING REACTOR | @ RE-FABRICATION
U: 30.3MT 4
Pu; 1.58 MT SPENT o MAKEUP FUEL
F.Po= LLI3MT 30.7MT U
OTHER ACTINIDES NAT. U)
RADIOACTIVE |g ’
WASTE DISPOSAL

U: 0.303 MT
Pu = 0.0158 MT
F.P.+ ACTINIDES - 1.13MT

POWER

FIGURE VII-B-7. Laser Hybrid Reactor - Pu-Recycle

VII-38
LITHIUM CRUDE ORES WATER THORIUM, URANIUM ORES
12248 MT 20789 MT U30g ORES
NAT. Li ORES 1658 MT ThOo ORES
Li MINING DEUTERIUM U, Th MINING
AND MILLING PROCESS ING AND MILLING
582 MT Li D: 19.8 kglyr 39.5 MIU
T: 29.7 kglyr 79 MT Th
EXTERNAL Pu:
14 MT
. FERTILE
Li ENRICHMENT BLANKET )
FABRICATION FABRICATION
' U: 39.1MT
103 MT Li%0 Th: T8 MT
(48 MT Li) Pu: 3.5 MT RECYCLE
> u
1 :
- SPENT Pu0;
Lig0 - 25.7 MT | LASER MODULES MODULE
o HYBRID REACTOR REPROCESSING [ —.
"
TRITIUM
RECYCLE
y
- BRED FUEL
SEPARATION OF RADIOACTIVE L-233
D, T, Li, He {35 kglyr WASTE DISPOSAL 2584 kg U-233/yr
T BRED
TRITILM 5.3 kg Thyr
B.R. - 118 |STORAGE l
Y
TRITILUM FOWER UNBURNED FUEL FISSILE FUEL
STORAGE 1567 MWe D, T, Li STORAGE
D: 376 kglyr
T: 564 kghyr 272 kg U-233/PWR

RADIOACTIVE

RECYCLE: 458 kg U-233

I

b

REPROCESSING

U. 7.4 MT
Pu: 105 kg
F.P.: +ACTINIDES

= 1053 kg
Th: 22 MT

WASTE DISPOSAL |

L= 0,074 MT
Pu = 0.001 MT
F.P. +ACTINIDES: 1053 kg

FIGURE VII-B-8.

LIGHT WATER LWR
REACTOR RE-FABRICATION
h
SPENT TOTAL FUEL
FUEL LOAD = 30.5 MT MAKEUP FUEL
Th: 22.1MT
U: 7.9MT
'
POWER
1300 MW,

VII-39

Laser Hybrid Reactor - Pu-Catalyst
TABLE VII-B-17. Theta-Pinch Hybrid Fuel Management Data

Number of Fuel Managment Regions 4

Blanket Lifetime 4 years
Plant Capacity Factor 0.75

Blanket Replacement time 35 days

Fuel management Interval 1.3 years
Maximum Blanket Exposure n 3-4 Mwy/m2
Number of Blanket Modules 200

Number of Blanket Modules Replaced 50

The initial loading of fertile fuel, structure, tritium breeder, and
graphite reflector is given in Table VII-B-18.

The 30-year plant lifetime fuel requirements, based on a four-year
burnup cycle, are presented in Tables VII-B-19 - VII-B-21. The fuel mass
flows for the Pu-Recycle and Pu-Catalyst fuel cycles are given in Figures
VII-B-9 and VII-B-10. The Theta-Pinch Hybrid can support 7-8 PHRs based on
2592 kg Pu-239 annual discharge. The Theta-Pinch Hybrid supplies the U-233
makeup requirements for 18-20 PWRs. The Theta-Pinch Hybrid operating in
the Refresh cycle mode requires more than 20 years of fuel exposure before
the Pu/U enrichment reaches 2-3%.

C. FACILITY REQUIREMENTS

The major facilities of the hybrid fuel cycles are analogous and in
some instances identical to the LWR fuel cycle facilities now in use. Blanket
module fabrication, operational wastes, spent fuel storage, and reprocessing
facilities will be characterized for each hybrid fuel cycle type. The current
status of LWR fuel cycle technology relevant to the hybrid fuel cycle will be
discussed.

1. Fuel Fabrication - Mainline Process Description

a. Summary Description of Overall Process

The reference fuel and blanket module fabrication facility performs chemi-
cal and mechanical operations in the manufacture of hybrid blanket modules.

VII-40
Ly-1IA

TABLE VII-B-18. Theta-Pinch Hybrid Reactor Initial Material Requirements

Reactor/Fuel Cycle

Theta-Pinch/Once-
Through

Theta-Pinch/Pu
Recycle to Thermal
Reactors

Theta-Pinch/Refresh

Theta-Pinch/Pu
Catalyst

Fuel

2595 MT UC

2595 MT UC
1911 MT UO2
827.5 MT UO2
2160 MT ThC
66 MT PuO2

Structure

1688 MT ss (@)

1688 MT SS
1688 MT SS
- 1688 MT SS

Tritium

Breeding Material

1473 MT LiZO(b)

1473 MT Li,0
1473 MT LiZO
1081 MT L120

(a) Stainless steel amount includes cladding and module structure.

(b) Tritium breeding material, Lip0, enriched to 90% 6Li.

(c) Graphite is for reflector region of module.

Other

699 M1 ¢

699 MT C
699 MT C
699 MT C
Zv-1IA

TABLE VII-B-19. Once-Through and Pu-Recycle Fuel Charge Data

FUEL MANAGEMENT CHARACTERISTICS

Reactor Type

Fuel Type

Theta-Pinch Hybrid Reactor

Né tural Uranium Carbide

Annual
Capacity Factor, % kg

Year

Heavy Metal,

OONOOT W N —

10

75
75
75
75

(a) Metric Tons

2,469,787
617,446
617,446
617,446

617,446

Reactor Fuel Charge Data

Thorium, Uranium, kg (a) Plutonium, kg
kg 33 24 235 236 ;e W 40 24 242
0 0 0 17787 0 2452 0 0 0 0
. . 4446 - 613 . . . .
. . . 4446 . 613 . . . .
. . 4446 . 613 . . . .
0 0 U 4446 0 613 0 0 0 0
EV-1IA

- JTABLE VII-B-20. Once-Through and Pu-Recycle Fuel

FUEL MANAGEMENT CHARACTERISTICS

Reactor Type Tokamak Hybrid

Fuel Type Natural Uranijum Carbide

Reactor Fuel Discharge Data
Heavy Metal, Thorium Uranium, kg Plutonium, kg
kg kg 233 234 235 236 238 239 240 241 242

—<
)
I
~

Discharge Data

Fission
Products, kg

Other Isotopics, kg

Pa-233  Np-237 Am-241  (m-242

611,278 0 .12 1 4012 605.913 639 6.4 .058 "0
606,397 . .22 3 3722 600,000 1270 18 .25
601,607 . .30 5. 3447 594,176 1893 36 .64
596,767 . .37 7 3186 588,308 2508 59 1.25

s . + e
N = = (N

WOONO U &P

30 596,767 0 .37 7.5 3186 588.308 2508 59 1.25 0

254
508
762
1016

1016

~0 452 ~0 0
876 .24 .
1288 72 .
1680 1.66 .01

0 1680  1.66 .01
by-T1A

TABLE VII-B-21, Pu-Catalyst Fuel Charge Data

FUEL MANAGEMENT CHARACTERISTICS

Reactor Type Theta-Pinch Hybrid Reactor

U02/Pu02 Convertor Region

Fuel Type
ThC Breeding Region

Reactor Fuel Charge Data
Annual Heavy Metal, Thorium, Uranium, k Plutonium, kg

(a)

Year Capacity Factor, % kg kg 233 234 235 236 238 239 240 241 242

O OSSN O P wWw PN —

75 2,885,240 2,053,000 0 0 5252 0 724,168 58,366 25,627 13,702 5125
75 721,310 513,250 . 1313 - 181,042 14,591 6,406 3,425 1281
75 721,310 513,250 . - 1313 - 181,042 14,591 6,406 3,425 1281
75 721,310 513,250 . - 1313 - 181,042 14,591 6,406 3,425 1281

- - - - - .

- - L) -

75 710,115 513,250 0 0 1313 o 181,042 8,091 3,552 1,839 710

(a) LWR Discharge Plutonium
LITHIUM CRUDE ORES WATER URANIUM ORES
174598 MT l 328341 MT U30g ORES
NMZUOfiSl ‘ ‘
Li MINING DEUTERIUM U MINING
AND MILLING PROCESSING AND MILLING
D: 24.8 kqlyr
8203 MT Ui T. 37.2kglyr 524 MTU
Li ENRICHMENT FERTILE —
FABRICATION BLANKET
-ABRICATION
1473 MT Lio0 RECYCLE U
(684 MT Li)
617.6 MTU
'L L SPENT
L0 - 368 MT THETA-PINCH MODULES MODULE |
HYBRID REACTOR REPROCESSING
—_—— —
TRITIUM U: 591 MT
F.P.. 1016 kglyr
ACTINIDES: 1647 kg
ACTIVATED S.S.:
SEPARATION OF | RADIOACTIVE 2592 kq Pu-
D, T, Li, He BRED T WASTE DI SPOSAL 239lyr
= 44.2 kglyr
TRITIHUM
B.p.allg | KT B
TRITIUM POWER UNBURNED FUEL FISSILE FUEL
STORAGE 45 MW, D, T, Li STORAGE
D: 471 kqlyr
T. 707 kglyr 334 kg Pu/PWR
RECYCLE: 514 kglyr
L_ gly _»
LIGHT WATER LWR
REPROCESSING REACTOR ¢ RE-FABRICATION
It,” 3?'538% SPENT TOTAL FUEL | MAKEUP FUEL
uol. FUEL LOAD = 33 MT
F.P. = 1.13MT
RADIOACTIVE | OTHER ACTINIDES
WASTE DISPOSAL [
U - 0.303 M7 l';&wnm
Pu = 0.0158 MT ¢

F.P.+ACTINIDES = 1.13 MT

FIGURE V

11-B-9.

VII-45

Theta-Pinch Hybrid Reactor -
Pu-Recycle to Thermal Reactor
LITHIUM CRUDE ORES

WATER

THORIUM, URANIUM ORES

123102 MT l L l 96985 MT U30g ORES
NAT. Li ORES 10918 MT ThO2 ORES
Li MINING DEUTERI UM U, Th, MINING
AND MILLING PROCESSING AND MILLING
6084 MT Li D: 21.8kgr 184.3 MTU
pool.e Xkgly 519 MT Th
EXTERNAL Pu
66 MT
Li ENRICHMENT FERTILE
FABRICATION BLANKET |4
U: 182.4 MT FABRICATION
| - Th: 513.4 MT
1081 MT Lix0 Pu: 16.5MT RECYCLE
(502 MT Li) U
Th
, l A 4 r SPENT Pu0z
| THETA-PINCH MODULE
| HYBRID REACTOR REPROCESSING joem
TRITIUM
RECYCLE
A
| BRED FUEL
SEPARATION OF '44kg]7yr RADIOACTIVE U-233
D, T, Li, He | pord” WASTE DISPOSAL | | 5066 kgiyr
TRITIUM 6.7 kg Tlyr TO
B.P.-1.18 STORAGE [ — : —
TRITIUM POWER UNBURNED FUEL FISSILE FUEL
STORAGE 1557 MW, D, T Li STORAGE
D: 471 kglyr
T. 707 kglyr }
RECYCLE: 458 kglyr 212 kg U-233/PWR
J ;’
LIGHT WATER | LWR
REPROCESSING REACTOR RE-FABR ICATION
U 7.4 MT TOTAL FUEL
Pu. 105kg 1053 kg LOAD: 30.5 MT _________MAKEUP FUEL
F.P. + ACTINIDES Th: 22.1 MT
Th: 22 MT U: 7.9 MT
RADIOACTIVE  |q
WASTE DISPOSAL
F.P.+ACTINIDES=1053 kg 1§gown§5v
Th: 0.22 MT e
U 0.074 MT
Pu 0.001 MT

FIGURE VII-B-10.

Pu-Catalyst

VII-46

Theta-Pinch Hybrid Reactor -
The facility receives natural uranium (U308) concentrates and enriched L120
(90% 6L1'). The natural uranium is to be the only radioactive material present
in the reference facility. Four basic operations are performed in the facility:
chemical conversion of the U308 to UO2 powder and then to a carbide (UC);
mechanical processing including preparation of UC pellets by cold pressing and
sintering, fabrication of stainless steel clad UC fuel rods (the tubes are
loaded with finished pellets, fitted with end plugs, and welded), and manufac-
ture of L120 tritium breeding pins; manufacturing of first wall and module
structures and placing finished fuel and tritium breeding rods into the module;
and recovery of uranium and L120 from off-specification and scrap material.

The finished blanket modules are to be shipped to appropriate hybrid commercial
reactors.

Description of Process Steps

e Chemical Conversion of U505 to UC

U308 concentrates are received from the mill facility. This
material is reacted with hydrogen to produce U02. UC is then produced
by oxide-carbon solution preparation. This fuel conversion is iden-
tical to the process used in the light water reactor industry.

e Blending and Packaging

The UC from the conversion processes is pulverized and then
collected into leaches for blending and acceptable UC is packaged in
canisters for transfer to the pellet manufacturing area. Rejected
UC is recycled back into the conversion step. L120 is purified and
collected into canisters for shipment to the pellet operations area.

e Pelletizing

UC from the conversion or scrap recovery area is received in
the pellet area where it is prepared for low pressure pressing.
After UC is densified in the slug processing operation, the slugs
are granulated and screened to obtain the proper size. At the pel-
leting station the granulated densified UC is pressed into pellets.
These pellets are passed through a sintering furnace and then placed
in a drying oven.

ViI-47
® Rod Loading and Finishing

Dried UC pellets are unloaded and the pellets manually loaded
into stainless steel rods. The top and bottom of the rods are welded
and sealed. Lizo powder will be loaded directly into the tritium
breeding pins. The L120 powder is compacted within the rods and then
sealed with end plugs.

® Module Assembly

The finished UC and LiZO rods are unloaded from their storage
racks and containers and are inserted into the module body. The
fabrication steps of the stainless steel module structure are depen-
dent upon the hybrid reactor type. Some of the modules will be
wedge-shaped and some cylindrical. Once the modules are assembled
and completed, they are shipped to a hybrid reactor site.

® Scrap Recovery

Scrap in various forms is sent to the recovery process operation

where it is handled on a batch basis. Scrap recovery in this reference

plant is relatively clean uranium-containing scrap that is amenable to

recovery of uranium of acceptable quality by a modest amount of pro-
cessing (i.e., without solvent extraction). Dirty scrap requiring
more processing is either packaged and stored for later processing
or shipped as waste.

Initial steps in scrap recovery involve concentration and conver-
sion of the scrap into forms that can be readily processed into U308
powder. The basic sequence of the scrap recovery process involves:
dissolution of solid forms in nitric acid, conversion to slurry,
dewatering the slurry form by wet mechanical separation, calcining
the resulting sludge in regular or controlled atmosphere furnaces,
and packaging and storing the resulting product. Some scrap does not
require processing through the entire sequence. Acceptable product
is recycled by returning it to the powder preparation step in the

VII-48
pellet area. Unacceptable product is transferred to the pH adjust-
ment station or the calcination station in the conversion area.
Solid waste is collected for disposal.

Liquid effluents held in the quarantine tanks are transferred
to a waste treatment building when they do not exceed the specified
release leveis. A typical fabrication facility outlay is shown
schematically in Figure VII-C-1.

Table VII-C-1 gives a summary of the isotopic, physical, and chemical
characteristics of the material present in the fabrication facility. The
average quantity of feed material that enters the facility is also given.
The tritium breeding pins will not require continuous repliacement but
instead can be recycled into the fresh module. The graphite reflector
region can aliso be reused in the hybrid module. The only nonfuel material
required for module fabrication is stainless steel used as the cladding.

2. QQ_Z/PUO2 Fuel Fabrication

In the Pu-Catalyst fuel cycle 233

U is bred in a ThC breeding region of the
blanket moduie. In order to enhance this process and improve neutron multi-
plication, PuO2 is added to the converter region. In this region the amount
of PuO2 is relatively constant, 8%, because the rate of Pu production is
approximately equal to the Pu burnup rate. Of course, the U02/Pu02 pins will
need to be re-clad due to neutron bombardment and radiation effects. It is
unrealistic to assume that the fuel pins (converter region) would retain their
cladding integrity for a large number of cycles. In the following discussion
the mixed oxide fabrication facility is characterized.

3. _QQ_Z/PUO2 Fuel Fabrication Mainline Process Description

a. Summary Description of Overall Process

The reference UOZ/PuO2 fuel fabrication facility receives UO2 (natural U)
and PuO2 powder. The UO2 and PuO2 are blended with recycled U02/Pu02 powders
from other process steps in the facility. The blended mixture is cold pressed
and sintered to yield U02/Pu02 pellets which are loaded into tubes to produce
convertor region hybrid type, stainless steel clad, U02/Pu02 fuel rods. The
welded and inspected fuel rods are shipped to the hybrid fuel fabrication
facility for incorporation into appropriate blanket modules.

VII-49
——— - —— — o

14
1 3 5
4 6 8 10 11 12
7|7 [ TRUCKWELL
13
2
‘i_EgENTRANCE
15
1 - Chemical Processing
2 - Chemical Laboratory Area
3 - Pelletizing
4 - Sintering Furnaces
, 6, 7 - Fertile Fuel Rod Loading
8, 9 - X-Ray
10, 11, 12 - Blanket Module Assembly Area
13 - L120 Rod Preparation and Loading
14 - UC, LiZO Recovery Area
15 - Office-Control Area
FIGURE VII-C-1. Fabrication Facility Layout

VII-50
TABLE VII-C-1. Once-Through and Pu-Recycle to Thermal Reactors Fuel
Fabrication Facility

Fuel Type - Uranium carbide fertile fuel rods clad with stainless
steel. Lithium oxide (90% OLi) clad with stainless steel
is tritium breeding material.

Type of Material - Contact-remote fabrication unnecessary with natural
Handling uranium as fuel feed.

Technology Status - Fertile fuel pin fabrication utilizes LWR technology.
Some modification of ex1%t1ng fabrication facilities will
be needed for Lip0 (90% ) pin fabrication.

Throughput (Range expected for commercial operation):

Tokamak Hybrid - 70 to 80 MT/yr-reactor(a)
Mirror Hybrid - 70 to 80 MT/yr-reactor
Laser Hybrid - 110 to 120 MT/yr-reactor

Theta-Pinch Hybrid - 600 to 620 MT/yr-reactor

Material Stream Characteristics
Fissile Isotopic

Physical Form Chemical Form Composition
Feed Yellowcake Concen- U308 0.72% 235U
trates
Product  Fertile Fuel Pins uc 0.72% 3%y
with Stainless Steel
Cladding
Waste Airborne Particles, U Contaminated 0.72% 235
Solid and Liquid Material

Operational Wastes

Piant Modification Feasibility/Proliferation Criteria:
Material Flow Change: low feasibility
Process Change: low feasibility

Proliferation Criteria: fabrication of natural uranium fuel entails Timited
proliferation risks.

(a) Annual throughput of natural uranium for fertile fuel pin fabrication.

VII-51
The reference U02/Pu02 fuel fabrication facility contains equipment in
"canyon" type areas where mixed oxide pellets are fabricated on a remote,
batch-type basis. The fabrication plants for the Pu-Catalyst hybrid fuel
cycle will use batch-type operations and semiremote (glove box) methods for
all the steps up through the final welding on the fuel rod. Mixed oxide fuel
is currently prepared commercially by dry (mechanical blending) or wet (copre-
cipitation) processes. The hybrid fuel fabrication facility will use the dry
technique in the pelletization process. (Figure VII-C-2.)

b. Description of Process Steps

Each of the major processes involved in manufacturing the U02/Pu02 rods
will be discussed below. This description will apply only to the U02/Pu02
portion of fuel fabrication for the Pu-Catalyst fuel cycle.

e Pu0, Receiving/Unloading

The PuO2 arrives in special shipping containers from a reprocessing
facility. The containers are opened inside a ventilated enclosure, the
inner cans opened, and the PuO2 transferred to a restricted storage
vessel.

e Powder Blending

Natural U02, Pu02, and recycled U02/Pu02 powder are blended in
batch increments. The UO2 powder is shipped from a fuel fabrication
facility to the U02/Pu02 facility in 55-gal drums. The U02/Pu02 mixture
is 92% (weight) Uo, and 8% PuO2 (weight). Rejected UO2 and PuO2 (71%
moisture) goes to scrap recovery and drying. The batch of U02/Pu02 is
transferred from the blender to a storage area.

e (Compaction, Granulation, Pelletization and Pellet Storage

U02/Pu02 powder is transferred by the air conveyor from a silo
to the slug press where the powder is compacted into slugs which are
then granulated and classified. Acceptable granules are sent to the
pelletizing press, oversize granules (indicates broken classifier
screen) are considered dirty scrap, and undersized granules are directly
recycled back to the slug press. Acceptable green (i.e., unsintered)
pellets from the pelletizing press are moved by mechanical conveyor to

VII-52
FEED MATERIAL AND
PER SONNEL CONTROL BUILDING ™

ROD REPAIR
AND DISMANTL

ELECTRICA; SUBSTATION

, J

ELECTRO-MECHANICAL BUILD ING

AN
N

SWITCHGEAR

LOCKER ROOM

T
[—

RY

J T

ING

p—

r————:flf

ROD INSPECTION
BUILDING

—!

U
I
|

7/| MEC HAN ICAL EQU | PMENT

FAN ROOM

ANALYTICAL SERVICES
FACILITY

|——— MANUFACTURING BUILDI . »

™~ DECONTAMINATION
AND HOT REPAIR CELL

—

“w]
™~ TRUCK WELL

FUTURE
ROD INSPECT
BUILDING

HVAC EXHAUST STACK
DIESEL EXHAUST STACK

DIESEL ROOM AIR INLET
STACK

TRUCK WELL

FEED MATERIAL AND
PERSONNEL CONTROL BUILDING

COLD CHEMICAL STORAGE

SINTERING FURNACE AREA

FUTURE

4——'— MANUFACTURING

BUILDING

" COOLING TOWER

AREA
ION _—— -
| ROD LOADING
— — —1 AND WELDING
FIRST LEVEL
ELECTRICAL SUBSTATION
4 1
ELECTRO-MECHANICAL
BUILDING L
.
L |

4,._--—"'-

FRESH AIR
INTAKE

-

MISCELLANEOUS

L WASTE TREATMENT

———"FILTER ROOM

MANUFACTURING
BUILDING

. POWDER STORAGE
AND SCRAP RECYCLE

Fabrication Facility for

UDz STORAGE 1L
AND UNLOAD ING —
1r—
FEED MATERIAL —
RECE IV ING AND STORAGE 5 }
(T e rclnlm | 712
| g
ROD INSPECTION
BUILDING -
{UPPER VOLUME) — pa
| FUTURE MANUFACTURING
FUTURE ROD | BUILDING
INSPECT ION BUILDING ‘
{UPPER VOLUME! | Puo, sToRAGE
I
_______ 0 U |
SECOND 1EVLL
FIGURE VII-C-2. U0 /PUO%
Pu- Cata yst Fuel Cycle

VII-53
boats (i.e., trays) which are, in turn, placed in green pellet storage;
rejected green pellets are collected as clean scrap.

Sintering, Boat Conveyance and Pellet Storage

Boats of green pellets are moved by shuttle car to the sintering
furnaces. The fabrication facility will apparently have several sin-
tering furnaces. The boats of pellets pass through the furnaces to an
inspection station for sintered pellets. Also, the acceptable pellets
are sent to sintered pellet boat storage, underfired pellets are
recycled back through the furnace, and overfired pellets are sent to
scrap recovery.

Pellet Grinding, Inspection and Storage

Boats of sintered pellets from the furnace storage area are trans-
ferred by motor-driven conveyor. The pellets are mechanically unloaded
and conveyed single file through a centerless grinder for surface grinding
(water coolant used). The water coolant used in pellet grinding is
pumped to a sludge separator. A high-velocity air stream (from nozzles)
passes over the ground pellets and dries them.

The ground pellets are inspected for diameter. Acceptable pellets
go to the nick inspection station, undersize pellets are transferred to
scrap recovery, and oversize pellets are sent back for regrinding. From
the nick inspection station, acceptable pellets are mechanically con-
veyed single-file to a tray loader. The loaded trays of pellets are
mechanically conveyed through a heated-air drier. Trays of dried,
inspected pellets are mechanically conveyed to the pellet storage unit.
Before pellets are released from the storage units for insertion into
tubes at the loading station, selected trays of pellets are conveyed
to the inspection and sampling station. Trays of acceptable pellets
are moved from the storage units to the loading station.

Fuel Rod Loading

The fuel rod loading station is in a glove box and has a mechanical
device to load peliets into stainless steel tubes. After the rods are

VII-54
loaded, they are removed to a decontamination station. From here, the
finished rods are sent to an inspection station where the welds and
dimensions of the rod are checked. Acceptable rods are moved to a
storage area before being shipped to the fuel fabrication plant where
the breeding pins, ThC, are manufactured. Here the UOZ/PUO2 rods are
inserted into the completed blanket modules.

Table VII-C-2 lists the fuel fabrication facility characteristics
for the Pu-Catalyst fuel cycle.

4. Refresh Fuel Cycle Fabrication

The Refresh fuel cycle employs a direct exchange of fuel between the
hybrid and 1ight water reactor. The initial fabrication will be identical to
the UO2 fuel pin fabrication for LWRs. Before the hybrid spent fuel can be
loaded into a LWR, the fuel is mechanically pressed out of the stainless steel
cladding. The fuel separated from the old cladding is transferred to another
area of the fabrication facility where it is baked to remove fission product
gases and then compressed and machined into pellets before it can be clad in
"Zircaloy" tubing. A1l of these operations must be performed on a remote basis.

5. ThC Fuel Fabrication

The Pu-Catalyst blanket module fabrication facility performs the chemical
and mechanical operations in the manufacture of hybrid fuel rods. The facility
receives natural uranium, thorium, and Pu02. The UOZ/PUO2 or convertor fuel
rod fabrication process was described previously. This section will charac-
terize the manufacturing of the breeder region fuel rods, ThC.

a. Chemical Conversion/Packaging

Thorium nitrate tetrahydrate (Th(N03)44H20) is received from the mill
facility. The thorium nitrate solution must first be converted into a finely
divided powder, Th02. The nitrate solution is transferred to a vessel where
superheated steam is used to drive off nitric acid leaving a ThO2 powder. The
oxide solution is then transferred to another area where the oxide is formed
into an oxide-carbon solution by heating it with channel-black carbon added to
the vessel. The ThC material is then collected and purified before being
transferred to the peliet process area.

VII-55
TABLE VII-C-2. Pu-Catalyst Fuel Fabrication Facility Characteristics

Fuel Type - Mixed oxide (U0O2/Pu0,) convertor fuel pins ctad with
.staintess steel. ThC and Lip0 (90% 6Li) breeding
material clad with stainless steel.

Type of Material - Remote fabrication processes carried out in a hot cell
Handling operations area. Initial Pu0; loading is LWR grade
ptutonium.

Technology Status - Mixed oxide convertor pin fabrication based on technology
developed for the LMFBR.

Throughput (Range expected for commercial operation):

Tokamak Hybrid - 20 to 30 MT/yr-reactor U
50 to 60 MT/yr-reactor Th
3 to 4 MT/yr-reactor Pu

Mirror Hybrid - 20 to 30 MT/yr-reactor U
45 to 55 MT/yr-reactor Th
3 to 4 MT/yr-reactor Pu

35 to 45 MT/yr-reactor U
75 to 85 MT/yr-reactor Th
5 to 6 MT/yr-reactor Pu

to 185 MT/yr-reactor U
500 to 600 MT/yr-reactor Th
20 to 30 MT/yr-reactor Pu

Laser Hybrid

Theta-Pinch Hybrid

1
—
os)
[an}

Material Stream Characteristics
Fissile Isotopic

Physical Form Chemical Form Composition
Feed Yellowcake Concen-  U30g, ThOp 724 £330 (V)
trates, Thorium and PuO2 %ZO% Pu and
Ores, and Plutonium 2%py  (Pu)
from LWR Reprocess-
ing
Product Convertor Fuel Pins  UOp/Puo, 724 530U (U)
and Fissile Fuel ThC , 70% Pu and
Breeding Pins 241py  (Pu)
Waste Airborne Particles, U Contaminated  .72% 23°U (b)
Solid and Liquid Material, Pu

Operational Wastes

Plant Modification Feasibility/Proliferation Criteria:
Material Flow Change: low feasibility
Process Change: tow-medium feasibility

Proliferation Criteria: fabrication of mixed oxide entails proliferation
risks. Diversion proof measures are required at
this facility.

VII-56
b. Pelletizing

ThC from the conversion or scrap recovery area is received in the pellet
area. After the ThC is densified in the slug processing operation, the slugs
.are granulated and screened to obtain the proper size. At the pelleting
station the granulated, densified ThC is pressed into pellets. These pellets
are passed through a sintering furnace and then placed in a drying oven.

The remainder of the fabrication process (rod loading and finishing,
module assembly, scrap recovery) are identicial to the processes for the other
fuel manufacturing facilities.

6. Hybrid Fuel Storage

Most LWR reactors have spent fuel storage pools that are capable of
handling 1-1/3 to 2 core loadings. The blanket module management plans for
the hybrid reactors are based on replacing one-fourth of the blanket each
year. The spent fuel storage for the hybrids should have a capacity of 1-1/4
of the total blanket loading. The spent fuel basin will utilize water as a
coolant and shield. Special storage canisters will be needed to store the
spent fuel rods since the module structure itself is reused in the blanket.
Additional storage may be required at the reactor for Once-Through fuel cycles.

7. Operational Waste Facilities

In addition to the wastes generated by blanket replacement operations
there are wastes resulting from the operation of the hybrid that are not pres-
ent in a fission reactor station, These wastes are associated with the
operation of a fusion reactor. Tritium contamination of the primary coolant
system occurs. There are also tritium wastes associated with the vacuum
system and cryopump systems. Recovery bed wastes (contaminated sieve beds)
will be generated by the bred tritium removal system. This system is used
to remove tritium bred in the L120 pins of the blanket module, From a prolif-
eration or diversion perspective, these types of wastes pose no risk.

8. Reprocessing - Spent Hybrid Fuel

There are two fuel cycles or blankets which will require reprocessing: Pu
Recycle to Thermal Reactors and Pu-Catalyst. The Once-Through employs a throw-
away blanket concept optimized for power production only. The Refresh fuel
cycles will use a mechanical type separation process.

VII-57
There are several options available in the hybrid spent fuel repro-
cessing which render the product less vulnerable to proliferation. The
Purex process can be adjusted to produce a coprocessed product, Pu and U.

Or the normal Purex process can be adjusted to yield coprocessed U, Pu with
fission product spike (partial decontamination). The Thorex process will be
used to reprocess Pu catalyst spent fuel. However, the U02/Pu02 portion will
usually remain inside the module and will not be replaced with fresh fuel.
The cladding will need to be replaced periodically.

9. Pu Recycle to Thermal Reactor Reprocessing: Mainline Process Descriptions

The mainline processes employed at the hybrid reprocessing facility can
be divided into three main categories. These are: 1) the process by which
the uranium and plutonium are recovered in highly purified nitrate solutions,
2) the process by which the purified uranium is converted from nitrate solu-
tion to uranium carbide, and 3) the process by which the purified plutonium
is converted from nitrate solution to plutonium dioxide.

The uranium nitrate solution is converted to a carbide form and recycled
back into the hybrid reactor. The PuO2 is transferred to a mixed oxide fab-
rication system where it can be used in light water reactors. An option that
is available in the reprocessing step is to leave the U and Pu in solution
(co-process) and use this fuel as light water feed material. The first pro-
cess description will apply to partitioned U/Pu streams. The significant
processes present in coprocessing will be emphasized following the partitioned
processing discussion.

10. Description of Process Steps

a. Recovery of Uranium and Plutonium

The hybrid reprocessing facility uses the Purex recovery process, which
has been in large scale use for over 20 years and is currently employed, with

minor variations, by most of the reprocessing plants now operating throughout
the world.

VII-58
e Spent Module Receiving

The irradiated hybrid fuel rods arrive at the facility in
shielded casks. The cask and carrier are monitored for outside radio-
active contamination to determine if any leakage has occurred and are
washed. The hybrid spent fuel elements are stored on reactor site in
a decay heat pool for approximately one year.

e Hybrid Fuel Rod Shearing and Dissolution

The fuel rods are remotely transferred from the storage pool to the
feed mechanism of the mechanical bundle type shear after a full pro-
cessing 1ot has been accumulated. Here the fuel elements are chopped
into segments about 5 to 12 c¢m long to expose the fuel to the dissolvent.
The fuel segments fall into the dissolver containing hot 3-8M nitric
acid (and gadolinium nitrate which serves as a neutron poison), which
dissolves virtually all the uranium, plutonium, and fission products.

The undissolved cladding materials and accompanying hardware of stain-
Tess steel remain in the dissolver basket. The dissolver solution is
centrifuged to remove fine solids which are sent to the high level

waste storage system. The clarified dissolver solution is transferred

to tanks to be sampled for accountability and to adjust the acid con-
centration to 2-3M nitric acid before being fed to the solvent extraction
process. The cladding huils are rinsed, monitored for residual fissile
material, packaged, and transferred to the interim underground waste
storage area.

e Solvent Extraction, Partitioning, and Stripping of Plutonium and Uranium

After acid adjustment, the feed solution is sent to the first solvent
extraction cycle where it is contacted countercurrently in a centrifugal
contractor with an organic solution of tributyl phosphate. The lighter
organic solution preferentially extracts the tetravalent plutonium and
hexavaient uranium, leaving about 98 percent of the fission products in
the aqueous solution. The organic solution from the centrifugal
contractor passes through a pulsed scrub column where a nitric acid
solution removes about 98 percent of the remaining fission products

VII-59
and is recycled back to the centrifugal contactor. The final aqueous
solution Teaving the centrifugal contactor contains about 99.9 percent
(or more) of the fission products, essentially all of the transplutonium
elements and about 0.5 percent of the uranium and plutonium; it is
then sent to a highlevel waste concentrator. The organic solution
from the pulsed scrub column then is joined by organic raffinates from
the plutonium purification sections and passes through a partitioning
column where tetravalent plutonium is electrochemically reduced to the
less extractable trivalent state. This enables the plutonium to be
stripped into an aqueous nitric acid solution containing hydrazine as
a holding chemical reductant, all within the same electrochemical
device. The organic solution passes through the final first cycle
pulsed column where the uranium is stripped into acidified water.

Second Uranium Solvent Extraction and Concentration

The aqueous strip solution containing the uranium is concentrated
adjusted with nitric acid and is sent to the second uranium solvent
extraction cycle where it is again preferentially extracted by another
organic solution in a pulsed column. Before leaving the column, the
organic solution is scrubbed with nitric acid solution which removes
additional fission products. Hydroxylamine nitrate and hydrazine are
also added to the scrub solution to remove residual plutonium by chemical
reduction to the less extractable trivalent state. Uranium is stripped
from the organic solution in another pulsed column, using acidified water.
This solution is concentrated further by evaporation. Finally, the
concentrated uranium solution from the second cycle is passed through
silica gel beds, if necessary, to remove residual traces.

Second Plutonium Solvent Extraction

Plutonium in the aqueous stream leaving the partitioning column in
the first cycle is reoxidized to the extractable tetravalent state
with nitrogen dioxide or sodium nitrite and sent to the second plutonium
solvent extraction cycle. Here it is preferentially extracted into an
organic solution in another pulsed extraction column. In the top portion
of the same column, the organic stream is scrubbed with nitric acid

VII-60
b.

solution to remove residual extracted fission products. The organic
Stream passes through a strip column where tetravalent plutonium is
transferred to an aqueous stream of dilute nitric acid.

Third Plutonium Solvent Extraction and Concentration

The extraction-scrubbing sequence is repeated in a third plutonium
cycle .for further decontamination from fission products. To effect a
higher plutonium product concentration, the plutonium is reduced in
the third strip column by hydroxylamine nitrate to the more readily
strippable trivalent state. A organic scrub solution is added to remove
residual uranium from the plutonium aqueous stream as it leaves the
third strip column. The plutonium product solution is analyzed and
stored in geometrically favorable tanks until it is transferred to a
facility for conversion to Pu02.

An overall analysis of the uranium and plutonium recovery process
shows that the uranium and plutonium product streams contain about one-part
in ten million of the fission products originally present in the spent
fuel. This purity translates to a radioactivity level in uranium of about
twice that of natural uranium. The radioactivity levels in the various
processing areas range from very high levels that require artificial
cooling to remove the heat from radioactive decay to levels low enough
to permit direct personal contact.

Conversion of Uranium Nitrate to Uranium Carbide

The fuel reprocessing facility also converts uranium nitrate solutions to

uranium carbide.

Uranyl Nitrate Receiving and Storage

The conversion area receives uranyl nitrate solution recovered from
spent fuel in the adjoining separations area. The solution is received
in an accountability tank where it is measured, sampled, and then trans-
ferred to the storage tanks.

VII-61
e Uranyl Nitrate Concentration

From storage, uranyl nitrate solution is pumped to a steam-heated
thermo-syphon reboiler where water is removed to form uranyl nitrate
hexahydrate (UNH), containing 78.5 weight percent uranyl nitrate. Removed
water is condensed and returned to the separations facility for recycle.

e Uranyl Nitrate Hexahydrate Calcination

Next, the UNH is calcined to uranium trioxide (U03) in a bed of UO3
fluidized by superheated steam at 315°C. A controlled discharge of UO3
is withdrawn from the bed and fed to the next process step. By denitrat-
ing in steam, the nitrate values are converted to nitric acid (HN03)
which is condensed and returned to the separations facility for recycle.

® Uranium Trioxide Reduction

Calcined UO3 is then put through a feed preparation step where it
is sized to a uniform particle size, activated by the addition of H2504
and is converted to uranium dioxide (U02) by reduction with hydrogen
in a fluidized bed, the hydrogen being obtained by dissociation of
ammonia.

The uranium dioxide produced by the reduction step is next reacted with
carbon in a furnace. After purification and further processing uranium
carbide is produced. This fertile fuel is stored and eventually recycled
back to the hybrid reactor.

¢c. Conversion of Plutonium Nitrate to the Dioxide

The hybrid reprocessing facility's plutonium production facility converts
plutonium nitrate solutions to plutonium dioxide powder.

The conversion process consists of continuous precipitation of plutonium
oxalate followed by calcination to plutonium dioxide (Pqu). This process
has been used for over 20 years in various nuclear installations. Two
parallel conversion lines (i.e., precipitation through product packaging)
are provided, each furnishing half the total capacity.

VII-62
e Plutonium Nitrate Conversion

Plutonium nitrate solution is transferred in batches from plutonium
nitrate storage to feed preparation tanks. In these tanks, the nitric
acid concentration is adjusted. The adjusted feed and oxalic acid
streams flow continuously to a precipitator vessel where they are mixed
and precipitation commences. From the precipitator vessel the slurry
overflows to successive digestion vessels to allow crystal growth. The
slurry is filtered on a rotary vacuum drum filter. The precipitate is
then dried and calcined in a rotary screw calciner at temperatures up
to 750°C. The plutonium oxide power is screened and blended to achieve
product uniformity. The oxide is then sampled, packaged and storaged
before shipment to a mixed-oxide LWR fuel fabrication facility.

11. Thorex Process for U/Th Reprocessing in the Pu-Catalyst Fuel Cycle

The Thorex process decontaminates uranium/thorium nitrate solutions and
separates it from the fission products. The mixture of nitrate solutions is
contacted with an organic solvent. The fission products are thus separated
from the uranium and thorium. The mixture is contacted in a second extraction
step in order to partition the uranium and thorium. These separate streams
are recycled through solvent extraction steps to remove the remainder of the
fission products. Purified uranium and thorium nitrate solutions are sent
to a recycle or refabrication facility.

12. Reprocessing Options

Listed below are the reprocessing facility options that exist for the
Pu-Recycle and Pu-Catalyst fuel cycles:

Process
Purex 1 Reference Purex process
2 Coprocessed U, Pu
3 Coprocessed U, Pu with fission-product spike,
i.e., only partially decontaminated
4 Coprocessed U, Pu, pre-irradiated before

shipment

VII-63
Thorex 1 Reference Thorex process
2 Coprocessed U, Th with fission-product spike
3 Partitioned products (U, Th, Pu)

4 Partitioned U, Thy Pu to waste with fission
products

5 Recycle 233U; denature in process; Pu, 235U to

waste with fission products

233 235

6 Recycle U; denature in situ; Pu, U to waste
with fission products
233, 235,,. ..
7/ Recycle u, U; denature in situ; Pu to waste

with fission products

A summary of the reprocessing facility data for the Pu-Recycle fuel cycle
is given in Table VII-C-3. The summary data for the Pu-Catalyst fueling option
is tabulated in Table VII-C-4.

Section VIII will deal with some of the proliferation resistant measures
that can be applied to hybrid fuel cycles.

VII-64
TABLE VII-C-3. Reprocessing Facility Summary Data for Pu-Recycle
Fuel Cycle

Fuel Type (feed):

Reprocessing Method:

Technology Status:

Maintenance:

Throughput (Sp

Tokamak Hybr
Mirror Hybri
Laser Hybrid
Theta-Pinch

Irradiated fuel rods (UC) containing actinides, fission
products and bred fuel (Pu). Reprocessing facility will
receive fuel rods in special shipping containers.

Purex and modified Purex for proliferation resistance
measures.

Reference Purex method well developed and in use in
several countries. Modified Purex methods have little
commercial basis.

Remote, hot cell operations.

ent Fuel Reprocessed in a Commercial Facility)
id
d

70 - 80 MT/year-reactor
70 - 75 MT/year-reactor
110 - 115 MT/year-reactor
500 - 600 MT/year-reactor

Hybrid

Throughput (Range Expected for Normal Commercial Operation}

Tokamak Hybrid - P

Mirror Hybrid - P

Product Waste
u: 1900-2000 kg/year-reactor 763 kg/year-reactor
(96% fissile) 204 kg/year-reactor
U: 71236 kg/year-reactor 0.088 kg/year-reactor
(0.54% fissile) 19.5 kg/year-reactor
719 kg/year-reactor
u: 800-900 kg/year-reactor 265 kg/year-reactor
(n96% fissile) 194 kg/year-reactor
U: 69100 kg/year-reactor 0.08 kg/year-reactor
(0.54% fissile) 8 kg/year-reactor
691 kg/year-reactor

VII-65

Pu
U
Laser Hybrid

Theta-Pinch Hybrid - Pu:

TABLE VII-C-3. (contd)
Product Waste
- Pu: 1300-1400 kg/year-reactor 518 kg/year-reactor F.P.
(96% fissile) 315 kg/year-reactor 23/Np
U: ~111120 kg/year-reactor 0.31 kg/year-reactor 24]Am

(.54% fissile)

Feed

Product

Waste

13 kg/year-reactor Pu

1000 kg/year-reactor U
2500-2600 kg/year-reactor 1016 kg/year-reactor F.P.
(96% fissile) 1680 kg/year-reactor 237Np
591490 kg/year-reactor 1.6 kg/year-reactor 24]Am
(.54% fissile) 5915 kg/year-reactor U

Material Stream Characteristics

Chemical/Physical
Form

A ]

Spent UC fuel rods containing
actinides, fission products,
activated structure (S.S.)

Partitioned stream of pluton-
jum nitrate (Pu(NO3)4) conver-
ted to Pu0, and UC (N03)2

SS cladding hulls, acidic high
level waste from the extraction
and concentrator steps. High
level solid wastes from initial
centrifugation process.

VII-66

Isotopics

gggu ~ 99.3%
236U - 0.54% : U

u - 0.16%
238, _ 1.159
239 .

Pu - 95%
240 6 Pu
24-IPU - 3- %

Pu 0.095%

Isotopics same as
above

A1l fission products and actinides
other than U or Pu are disposed of as
wastes. Fission products are 1-2% of
initial spent fuel feed while acti-
nides other than U and Pu are ~0.6% of
initial spent fuel. Some of the more
important actinides are:

237Np - 0.56% of spent fuel
24]Am - 1.8 x 10'3% of spent fuel
24200 _ 1.6 x 10°°% of spent fuel
TABLE VII-C-3. (contd)

Plant Modification Feasibility/Proliferation Criteria
Material Flow Change: medium-high feasibility
Process Change: 1low feasibility
Proliferation Criteria: reprocessing facility located in a secure nuclear
center would present limited proliferation risks.
Technical fixes such as co-processing and fission

product spiking could also be employed as diversion
resistant measures.

VII-67
TABLE VII-C-4. Reprocessing Facility Summary Data for Pu-Catalyst

Fuel Cycle
Fuel Type (feed): Irradiated converter fuel rods SUOZ/PUO ) and breeder rods
(ThC) and bred fissile fuel (23 /EuO rods will be

re-clad and not chemically treated

Reprocessing Method: Thorex and modified Thorex for proliferation resistant

product.

Technology Status: Reference Thorex process has seen limited commercial
use.

Maintenance: Remote handling devices required heavy shielding - hot

cell operations necessary.

Throughput (Spent Fuel Reprocessed in a Commercial Facility -
Breeding Region Only)

Tokamak Hybrid
Mirror Hybrid

50 - 60 MT/year-reactor
50 - 60 MT/year-reactor
/0 - 80 MT/year-reactor
510 - 515 MT/year-reactor

Laser Hybrid
Theta-Pinch Hybrid

Throughput (Range Expected for Normal Commercial Operation)

Product Waste
3800-4000 kg/year-reactor 1648 kg/year-reactorlF.P.
Th: 50.000-55,000 kg/year-reactor 38 kg/year-reactor
500 kg/year-reactor Th

i
e

Tokamak Hybrid

Mirror Hybrid - U: 1500-1600 kg/year-reactor 663 kg/year-reactor F.P.
Th: 48,000-53,000 kg/year-reactor 15 kg/year-reactor 233U

500 kg/year-reactor Th
Laser Hybrid - 233y:  2500-3000 kg/year-reactor 1250 kg/year-reactor F.P.
Th: 75,000-80,000 kg/year-reactor 26 kg/year-reactor 233U

750 kg/year-~reactor Th
Theta-Pinch Hybrid - 233U: 5000-6000 kg/year-reactor 2146 kg/year-reactor F.P.
Th: 510,000-515,000 50 kg/year-reactor U

kg/year-reactor 5100 kg/year-reactor Th

- VII-68
TABLE VII-C-4. (contd)

Material Stream Characteristics

Chemical/Physical
Form _ Isotopics
Feed Spent U0 /PuO converter rods 233U - 0.72%
and ThC greed1ng pins contain- 234U - 98.2% U
ing actinides, fission products, U - 0.89%
activated structure (S.S.)
Product Partitioned stream of U and Th; Isotopics same as

converter region of U0y/Pu0p is  above
returned to fuel fabr1cat1on

facility.

Wastes SS cladding hulls acidic high A1l fission products
level waste from the extrac- and actinides other
tion and concentrator steps. than U are disposed
High level solid wastes from of as wastes. Fission

initial centrifugation process. products are 1-2% of
initial spent fuel
feed.

Plant Modification Feasibility/Proliferation Criteria

Material Flow Change: medium-high feasibility
Process Change: 1low feasibility

Proliferation Criteria: reprocessing facility located in a secure nuclear
center would present limited proliferation risks.
Technical fixes such as co-processing and fission
product spiking could also be employed as diversion
resistant measures.

VII- 69
D.

SECTION VII REFERENCES

R. V. Laney, and P. R. Huebotter (ANL), "Nonproliferation Criteria for
Nuclear Fuel Cycles," TANSA 28:320 (1978).

The Energy Daily. 6(142):1, July 25, 1978.

Proceedings of the Fifth Energy Technology Conference, Government
Institutes, Inc., Washington, D.C., April 1978,

e Proliferation-Resistant Nuclear Technology, Chauncey Starr,
President, Electric Power Research Institute, p. 103.

e Precedents for Diversion-Resistant Nuclear Fuel Cycles,
Floyd L. Culler, Jr., Executive Vice President, Electric Power
Research Institute, p. 111,

e A Fast Breeder System Concept, Milton Levenson, Director, Nuclear
Power Division; Edwin Zebroski, Director, Systems and Materials
Department, Nuclear Power Division, Electric Power Research
Institute, p.230

e Possible Long Term Options for the Fast Reactor Plutonium Fuel
Cycle, R. H. Flowers, K. D. B. Johnson, J. H. Miles, R. K. Webster,
United Kingdom Atomic Energy Authority, p. 256.

B. R. Leonard, Jr. and U. P. Jenquin, "The Quality of Fissile Fuel Bred

in a Fusion Reactor Blanket." Second Topical Meeting on the Technology

of Controlled Nuclear Fusion, Richland, WA, September 21-23, 1976. ERDA
Report CONF--760935, p. 2, 711.

M. J. Bell, ORIGEN - The ORNL Isotope Generation and Depletion Code.
ORNL-4628, Oak Ridge National Laboratory, Oak Ridge, TN, May 1973.

T. M. Helm, et al., Reactor Design Characteristics and Fuel Inventory
Data, Vol. 1. Hanford Engineering Development Laboratories, Richland,
WA, September 1977.

VII-70
VIII. PROLIFERATION RESISTANCE CONSIDERATIONS

A. INTRODUCTION - GENERAL CONSIDERATIQONS

The relevance of nuclear power programs to proliferation risk arises
mainly from the possibility that the potential access these programs may
provide to weapon-usable fissile material may influence either the decision
to seek nuclear weapons or the ability to implement such a decision.

The prevention of proliferation will not be assured by unilaterally
developing in the United States alternative fuel cyclies or delaying reprocess-
ing or the fusion-fission reactor with a uranium-plutonium fuel cycle. The
potential for further world-wide proliferation is both immediate and diffuse,
since there are over 200 commercial nuclear power reactors and at least as
many research reactors around the world producing plutonium today. Fusion-
fission reactors containing uranium are simply another potential source of
plutonium, whose use would increase the amount of plutonium which could be
reprocessed.

A distinction must be made between two kinds of proliferation that con-
cern today's policy makers. The first kind is a country-specific scenario
of nations close to weapon capability now: the near-term proliferation pro-
blem. It is this problem that must be dealt with on a case-by-case basis.
The second kind of proliferation, the longer-term problem, relates to the
world-wide advancement in nuclear and other industrial technologies: a more
general and abstract problem, but nonetheless real. It is this second kind
which forms the basis for reevaluating alternative fuel cycles by attempting
to control the role of plutonium in future nuclear power. The most difficult
aspect of this approach is that it is discriminatory: the problem becomes
one of defining those "qualified" (for using plutonium) without antagonizing

(1)

others.

As important as it is, the issue of terrorism and other forms of sub-
national diversion or theft of nuclear material are not defined as prolifera-
tion in this report. The distinction is not an artificial or formal one:
terrorist threats to nuclear material are of a different nature and are

VI1I-1]
susceptible to very different forms of protection than are the risks of
governmental diversion and national proliferation. Furthermore, governments
possess both resources and nearly uniimited authority and power to counter
subnational threats, while the risks of national diversion must be dealt with

through the relatively Timited tools of diplomacy, international institutions,
and sanctions.

1. The Issue of Reprocessing

It is important to keep in mind that there are many alternative routes
to nuclear weapons other than the acquisition of fissile fuel from a civilian
nuclear power reactor.(z) At the present time, plutonium separation in a
chemical reprocessing facility is regarded as a basic point of cohnection
between nuclear power and nuclear weapons capabi]ities.(]) If stockpiles of
plutonium were to accumulate in national hands, international safequards as
a means of detecting diversion, and therefore of deterring it by providing
advance warning, become less meaningful.

Reprocessing, however, is viewed in some countries as essential to the
prudent long-term management of nuclear waste, and there is reluctance abroad
to proceed with the large-scale exploitation of nuclear power until the means

for permanent waste management are in hand.(])

In some instances government
regulations require reprocessing and/or firm plans for waste management as a
pre-condition of installing additional nuclear power plants. Failure to
reprocess and to recycle recovered plutonium also will lead to the accumula-
tion of large quantities of spent fuel in many places; this accumulation

could represent both a hazard to public health and an increasing proliferation

risk in its own right.

Because hybrid reactors couid produce power as well as fuel to extend
the fuel supply for fission reactors, they are capable of fueling muitiple
burner-converters and can serve a useful function in the perceived market

5)

hybrid breeders must produce and sell power at least sufficient to offset

place by the year 2000. However, previous studies(3' conclude that

the power consumed by the devices in order to compete in the market place.
The sale of fissile material probably requires chemical processing of the
blanket to recover the fuel, although recycle without reprocessing may be

(6)

possible.

VIII-2
2. Fusion-Fission Reactors Studied

As explained in Section III, the four fusion drivers considered are the
tokamak, mirror, theta pinch, and laser inertial systems. Based upon the
state-of-the-art of existing plant design, no discernible proliferation
advantages could be identified for one driver system over another provided
all plants were normalized to the same amount of fissile fuel produced annually
and to the same location. As was shown in Section VII, however, different
drivers oroduce different amounts of plutonium.

The principal factor of fusion-fission systems which influence non-
proliferation considerations is the fuel cycle selected. The fuel cycle
options of Section VII were:

e No chemical reprocessing, with the options of

® Once-through throwaway/stowaway
e Refreshing and mechanical reprocessing

e Chemical reprocessing and recycling, with the options of

® Pu recycle

233

e Pu and U recycle

The systems without reprocessing will be considered in Section VIII-B and
those with reprocessing will be covered in Section VIII-C.

3. Fuel Cycle Operations of Interest for Non-Proliferation

The fuel cycle operations which are of interest are those which give
rise to the prospective availability for diversion of fissile materials
to illicit uses. They are: reprocessing which produces highly enriched
235 239 233 233U and

Pu, or
235U; and transportation and storage of spent fuel containing
233

235y and/or plutonium, and/or U. These operations are in

u, or U; processing and storage of plutonium,
highly enriched
highly enriched
practice generally separable, and in actual practice separated. Some may be

amenable to being co-located with others, while others may not be.

The fuel cycle operations which are not the subject of interest are:

uranium exploration, mining and milling; conversion and fabrication of low

VIII-3
enrichment 235

U into fuel elements, and their transportation and storage as
"fresh" fuel elements even though low-enriched fuels may offer some improved
prolification resistance due to their diminished fissile fuel streams in the

conversion process.

4. Standard of Comparison

In order to place the hybrid concept in perspective it it useful
to relate the candidate hybrid fuel cycles to the fuel cycle scenarios
and technical fixes being considered for fission reactors. Such a per-
spective gives an indication as to whether these fuel cycles possess
desirable nonproliferation qualities which may permit the appropriate
criteria for proliferation resistance to be achieved. 3

B. NO REPROCESSING

These hybrid blanket concepts are discussed in Section VII where a com-
parison of the average U3O8 feed requirements and plutonium discharge per
year is tabulated. Compared with the LWR once-through system, these hybrid
blanket concepts offer greater proliferation resistance owing to the absence
of enrichment requirements, assuming that similar safeguards are provided
for the spent fuel. They also can have markedly improved resource utiliza-
tion since they can utilize depleted uranium or thorium. However, they
appear to be economically inferior since they involve plants with significantly

(5)

greater capital costs.

The second hybrid fuel cycle that operates in the no reprocessing mode
is the "refresh cycle" which is dissussed in Section VI. Their average
U3O8 feed requirements and plutonium discharged per year for the different
drivers are tabulated in Section VII.

In addition to the "refresh" cycle just discussed, any hybrid might be
used to "refresh" or re-enrich" normal spent fuel where the fresh fuel is
enriched to ~1.0% 235U in U at the end of its life. In this concept fission
reactor spent fuel would be shipped from the reactor discharge basin to a
refabrication center. The spent fuel would be mechanically refabricated
into fresh hybrid blanket module assemblies. This fuel would then be
re-enriched in the hybrid and, after an appropriate decay period, returned
to the fission reactor.

VIII-4
Conclusion about no reprocessing with fusion-fission hybrid reactors:

With no reprocessing, the principal advantage of hybrids (viz.. their ability
to produce copious amounts of fissile fuel) is lost.

C. REPROCESSING AND RECYCLING

These hybrids are somewhat analogous to fission breeders in that they
extend natural resources by converting uranium and/or thorium to fissile
material. The applications include hybrid blankets which produce only fissile
material for sale to support fission reactors as well as those which produce
both fissile material and electricity (or synthetic fuel) as salable producté.
Variants on the blanket fuel cycle include use of uranium, thorium, or
mixtures of both.

1. Plutonium Recover and Recycle

Since LWRs do not convert a sufficient amount of plutonium to completely
fuel themselves, an external source of plutonium is needed to sustain the
system and allow it to grow. In this case, the hybrid could be the external
source of proliferation resistant plutonium. The sources of uranium include:
mixed natural, depleted uranium from the enrichment plants and/or the uranium
recovered in reprocessing spent UO2 LWR fuel. Material flows for these
plutonium-recirculating cycles are tabulated in Section VII.

For concepts involving recycling of plutonium to fission reactors, pro-
liferation resistance may be adequate only if the hybrid, reprocessing and
fuel refabrication facilities are located in a secure International Nuclear
Center (INC) and

® The fissile and fertile materials are kept together at all times (e.qg.,
co-processed U-Pu) to dilute the fissile content to below weapons-grade,
or

e The fuel is made highly radiocactive (e.g., having highly radiocactive
materials in the fuel) to preclude handling, or

(7)

® The above two are combined in the CIVEX process.

VIII-5
It also has been proposed by Allied-General Nuclear Services that
denaturing plutonium by mixing it with a sufficient quantity of 238Pu can
make the plutonium unusable for weapons because of its high heat generation
rate.(g) This could be accomplished by recovering uranium and 237Np from the
spent fuel. The isotope 236U builds up during irradiation of the fresh UO2
fuel in the LWR. Subsequent irradiation of the 250U and 237 238

Np produce Pu
in the plutonium.

233

2. Denatured U Cycle

The hybrid thorium cycle is described in Section VII where the U308
requirements for the various fusion drivers are tabulated.

Thorium blanket concepts which involve the recycling of 233

U denatured
fuels to fission reactors are more proliferation resistant than plutonium
recycling blankets even though they may also require locating the hybrid
reactor and reprocessing and refabrication activities in an INC. In this
case, the fission reactor fuel probably should be denatured by mixing with
238U. This concept has high resource utilization since it makes use of
thorium and recycled 233U which can produce relatively high conversion ratios
232} buitds up

in these cycles to the point where the radiation levels are sufficient to

in the thermal fission reactors. Furthermore, the isotope

require massive shielding during handling and processing. The requirements
for shielding are perceived as adding proliferation resistance to the fuel
cycle.

3. High Gain Mixed Cycle

A potentially more attractive hybrid blanket is one in which depleted
uranium and recycled plutonium are used for neutron and energy multiplication

in which 233U is then bred in a thorium region.(g)

238U and 239

producing 233U, a superior fuel for thermal reactors. In this cycle, depicted

Such a design incorporates

the superior performance of Pu in a high-energy spectrum while

in Figure VII-B-3, the hybrid, reprocessing, and refabrication plants should

all be within the INC and the plutonium is separated and sent to storage while
the 233U is used to feed the LWRs located outside the INC. The circulating

VIII-6
uranium is denatured with 238U. The U308 requirements and the buildup of

plutonium are also shown in Figure VII-B-3,

Conclusion on Pu recovery and recycle with fusion fuel and hybrid
reactions: With INC in which hybrid reactors and reprocessing facilities
could be located, hybrids have a great advantage because of their fissile
fuel production.

D. PROLIFERATION RESISTANCE ENGINEERING

1. Allowable Activities

There is no established technical fix which can be applied to the
fusion-fission reactor with a U-Pu fuel cycle to sever its potential 1link
with proliferation. However, it must be remembered that no technical fix
exists even for the 1ight water reactor program currently underway in the
United States since plutonium must be continually stored. The closest approx-
imation to a "technical fix" is to avoid chemical reprocessing of spent fuel
so that all countries, including the United States, are at least one step
removed from a ready supply of plutonium. This would be a continuation of
the "status quo” and can ultimately lead to shortage of fissile fuel and to
239Pu with 238Pu
Np in LWR fuel has yet to be technicd]]y established. If this con-

severe economic penalties. The possibility of denaturing
237
and
cept becomes technically viable, then hybrids would have a marked prolifera-
tion advantage since they are capable of producing copious quantities of

238 237 239Pu.

Pu and Np along with

In principle, any such measures or "fixes" which may be available to
fission reactor fuel cycles can also be employed in the hybrid reactor sys-
tem. Thus the hybrid reactor designs will be reviewed in the context of
the progress made toward making fission systems more proliferation resistant.

There are two basic approaches for enhancing proliferation resistance:
technical barriers to proliferation (e.g., the isotopic, radiation, or repro-
cessing access barriers) and institutional barriers (e.g., special siting
constraints for sensitive fuel cycle facilities and storage of sensitive
fissile materials). As pointed out above, the proliferation resistance for
hybrids should be accomplished with a combination of the two approaches:

VIII-7
e the hybrid reactor, any reprocessing and refabrication facilities, and all
storage facilities should be located within an International Nuclear Center(]o)

e some form of radiation barrier, denaturing and/or co-processing barrier, or
both should be used for all fissile material recycled to the LWRs.

In addition,(1]) no plutonium should be permitted outside the INC except in
full reactor subassemblies, containing fission product activity to a level of
>1000 rem/h at 1 m. Furthermore, no such subassembly should leave the fuel cycle
center unless the plutonium content plus any other fissile material present in
the fuel, is <W% by weight in the fuel, where W = 30 for oxide fuel, 20 for
metal fuel, and 35 for carbide and nitride fuel.

No 233U shall Teave the INC except when contained in full reactor sub-

assemblies. A radiation level of >1000 rem/h at 1 m is required when the 233y
content in uranium exceeds X%m where X = 4. The upper limit of 233
fissile material in the fuel is W%, where W is defined as above.

235

U plus other

In the case of 235

U, bulk material may leave the INC provided the U
content in uranium is <4%. Higher levels require fabrication into full sub-
assemblies within the center. Gamma activity >1000 rem/h at 1 m is required
when the 235U concentration in uranium exceeds Y%, where Y = 12. The upper
limit of 235U plus other fissile material in the fuel is Z%, where Z = 40 for

oxide fuel, 30 for metal fuel, and 35 for carbide and nitride fuels.

A11 spent subassemblies containing chemically separable fissile material
should be returned to a fuel cycle center before the activity level drops below
1000 rem/h at 1 m.

The intent of these last four restrictions on form and condition of fissile
material outside of fuel cycle centers is to introduce no temptations for diver-
sions of material at either the front or back end of the once-through LWR cycle.
Upper 1imits are set on the fissile content of fuel to prevent the direct con-
version of a stolen subassembly to a weapon with only simple mechanical operations
such as duct removal and fuel pin chopping.

VIII-8
Co-processing with conventional solvent extraction systems can produce
about 12% Pu and subsequent conversion of the mixture to a form suitable for
mixed-oxide fuel fabrication seems technically feasible but has not been demon-
strated on a commercial sca]e(]z). However, with regard to nonproliferation,
plutonium can be recovered from co-processed material by simple chemical pro-
cesses (ion exchange, solvent extraction), so some form of radiation barrier

should also be added.

This addition of a radiation source to the uranium-plutonium mixture to
discourage unauthorized use can be accomplished by several techniques, as shown
in Table VIII-D-1. For proliferation purposes, no added advantage can be asso-
ciated with spiking beyond that required to assure the need for shielded equip-
ment to process the plutonium mixture.

The promising fission product candidates for spiking also have been iden-
tified, but no isotope has the ideal combination of properties. Cobalt, cesium
(provided it can be made nonvolatile), and/or cerium are the most promising

candidates(]z).

TABLE VIII-D-1. Methods of Spiking Plutonium' %)

Spikihg
Effect Fuel
Method Fabrication Means to Defeat Radiation Level

With Fission Products

Incomplete Removal Yes Recycle Depends on design

Selective Partition Yes Avoid add-back High decay depends on

and Add-Back isotope

Irradiate Fuel After NG Bypass Adjustable, will decay

Fabrication irradiation faster
With Cobalt Sources

Mix with Pu Fuel Yes Separate High decays with 5-yr.

half-1ife

Attach source to No Remove source

Fuel Assembly at

Pu Fabrication Plant
With 238Pu or 236Pu No Isotope Minor (~few R/h)

separation

VIII-9
2. Proliferation Resistance Effectiveness Evaluation

Proliferation resistance effectiveness evaluation (PREE) is the process of
estimating how effective such measures are at any given site or transportation
link. As yet, there is no established historical record from which to evaluate
proliferation resistance, and so predictive models are needed.

One approach to predictive proliferation resistance evaluation can involve
determining the probabilities of proliferation prevention for specified pro-
liferation scenarios. Such overall probabilities involve products of proba-
bilities of (a) a nation having the resources to attempt a diversion of special
nuclear material in order to fabricate a nuclear weapon, (b) a nation attempting
such a diversion, and (c) another nation (and ultimately the United States) not
detecting such a diversion with sufficient advanced warning that diplomatic nego-
tiations-would fail.

A number of scenario models have been developed for the domestic safe-

(]3']5), and it is not far-fetched to

guarding of special nuclear materials
envision some useful results from such techniques for assessing proliferation
as well. Unfortunately, these techniques lack sufficient detail at this time
to make meaningful comparisons between fusion-fission systems and LWRs, or
between different fusion-fission drivers. Some progress should be possible,

however, by examining the different fuel cycles in a generic manner.

VIII-10
E. REFERENCES FOR SECTION VIII

1. The Atlantic Council of the United States, Nuclear Power and Nuclear
Weapons Proliferation, Vol. I. Westview Press, Boulder, CO, 1978.

2. C. Starr, and E. Zebroski, "Nuclear Power and Weapons Proliferation,"
Proceedings of the American Power Conference, Chicago, I1linois,
April 18-20, 1977 and C. Starr, "Nuclear Power and Weapons Proliferation
- The Thin Link," Nuclear News, 20(6):55 (June 1977).

3. D. E. Deonigi, and R. L. Engel, Performance Targets for Fusion Fission
Hybrid Reactors BNWL-2139, Pacific Northwest Laboratory, Richland, WA,
1977.

4. D. E. Deonigi, "Economic Regimes," Proc. of the Second Fusion-Fission
Energy Systems Review Meeting Vol. II, CONF-/71155, Washington, D.C.
pp 327, 1978.

5. B. Augenstein, "Fusion-Fission Hybrid Breeders - Economic and Performance
Issues - Role of Advanced Converters, Interdependence Between Fission and
Fusion Programs." Proc. of the Second Fusion-Fission Energy Systems
Review Meeting Vol. II, CONF-771155, Washington, D. C., pp 297 (1978).

6. K. R. Schultz, R. H. Brogli, G. R. Hopkins, G. W. Shirley, and
S. C. Burnett, "Preliminary Evaluation of a U-233 Fusion-Fission Power
System Without Reprocessing," Proc. of the Second Fusion-Fission Energy
?ysteTs Review Meeting Vol. I, CONF-771155, Washington, D. C., pp. 183
1978} .

7. Proceedings of the Fifth Energy Technology Conference, ("Energy
Technology V, Challenges to Technology, February 27 - March 1, 1978,
Washington, D.C.). Government Institute, Inc., Washington, D.C., April
1978.

e Proliferation-Resistant Nuclear Technology, Chauncey Starr,
President, Electric Power Research Institute, pp. 103.

e Precedents for Diversion-Resistant Nuclear Fuel Cycles,
Floyd L. Culler, Jr., Executive Vice President, Electric Power
Research Institute, pp. 111.

e A Fast Breeder System Concept, Milton Levenson, Director, Nuclear
Power Division; Edwin Zebroski, Director, Systems and Materials
Department, Nuclear Power Division, Electric Power Research
Institute, pp. 230.

VIII-11
10.

11.

12.

13.

14.

15,

e Possible Long Term Options for the Fast Reactor Plutonium Fuel
Cycle, R. H. Flowers, K. D. B. Johnson, J. H. Miles, R. K. Webster,
United Kingdom Atomic Energy Authority, pp. 256.

The Energy Daily, 6(142):1, July 25, 1978.

S. F. Su, G. L. Woodruff, and N. H. McCormick, "A High-Gain
Fusion-Fission Reactor for Producing Uranium-233," Nucl. Tech. 29:392,
1976.

S. Strauch, "ATternative Reactor Fuel Cycles under Consideration and
Their Design Ramifications," Trans. Am. Nucl. Soc. 28:573, 1978,

R. V. Laney, and P. R. Huebotter, "Nonproliferation Criteria for Nuclear
Fuel Cycles," Trans. Am. Nucl. Soc. 28:320, 1978.

F. R. Field, and F. E. Touper, "Reprocessing and Fuel Fabrication
Systems," Trans. Am. Nucl. Soc. 28:67, 1978.

L. Kull, L. Harris, Jr., and J. Clancy, "A Method for Evaluating the
Effectiveness of a Facility Safeqguards System," Nuclear Materials
Management 6, 1977.

H. Kendrick, E. Loftgren, D. E. Rundquist, and R. R. Fullwood, "An
Approach to the Evaluation of Safeguards System Effectiveness," Nuclear
Materials Management 5, 1976.

H. A. Bennett, D. D. Boozer, L. D. Chapman, S. C. Daniel, B. C. Hulme,
and G. B. Varnado, "Safeguard System Effectiveness Modeling,” Nuclear
Materials Management 5, 1976.

VIII-12
IX. ECONOMICS

An assessment of the costs of constructing and operating the fusion-
fission (hybrid) power reactor and fuel cycle concepts introduced in the
technical sections ¢ this report is contained in this chapter. Estimates
of fusion-fission reactor system capital investment costs, operating and
maintenance costs, and fuel cycle costs are developed along with projections
of the resulting levelized energy costs or unit electricity costs. Also
developed are estimates of the break-even fissile values (i.e., the projected
selling value of the fissile fuel material produced in a commercial fusion-
fission reactor system). Projections of the extent to which the systems will
commercially deploy are given along with estimates of the economic benefits
that would accrue due to this deployment. An appraisal of the specific
economic penalties of utilizing proliferation resistant devices and fuel
cycles is also made.

A. GROUND RULES AND ASSUMPTIONS

The ground rules and assumptions utilized in estimating fusion-fission
reactor system costs are given in Table IX-A-1. These parameters were speci-
fied to ensure consistency in all phases of the evaluations.

B. CAPITAL INVESTMENT COSTS (Hybrid Reactor)

Capital investment cost or plant cost is the total cost of construct-
ing the hybrid reactor and placing the reactor into operation. Estimated
fusion-fission reactor capital investment costs for each of the reactor
driver/blanket combinations identified in this study are given in
Tables IX-B-1 and -2. Detailed cost estimates are given in Appendix A.

Estimates of reactor capital investment costs were generated assuming
the reactor is a commercial generating unit of optimal economic size. The
systems were costed assuming a mature industry (i.e., fifth facility of a
like technology constructed), thereby excluding development and "first of a
kind" costs from estimates. The following cost items or activities are
excluded from estimates.

IX-1
TABLE IX-A-1. Economic Parameters/Unit Costs

General Economic Conditions

Rate of General Inflation 0
Escalation Rate for Capital Investment Costs 0
Escalation Rate for Operating & Maintenance Costs 0
Escalation Rate for Fuel Cycle Costs 0
Base Year for Constant Dollar Analysis 1

System Description Data

Assumed First Year of System Construction 1978
System Operating Lifetime 30 years
System Construction Period 8 years

Utility Description Data

Annual "Other Taxes"

Annual Insurance Premiums

Effective Income Tax Rate

Ratio of Debt to Total Capitalization

Ratio of Common Stock to Total Capitalization
Ratio of Preferred Stock to Total Capitalization
Annual Rate of Return on Debt (Deflated)

Annual Rate of Return on Common Stock (Deflated)
Annual Rate of Return on Preferred Stock (Deflated)

OCOO0OO0OOOCOO0OO0O
£
o

Fission Fuel Cycle Unit Costs

Cost of Fertile Material (UC) $120/kg Heavy Metal
Cost of Fertile Material (UO,) $100/kg Heavy Metal
Cost of Fertile Material (DeB]eted Uranium) $7/kg

Cost of Fertile Material (ThC) $55/kg Heavy Metal

Cost of Fertile Material (ThO,) $35/kg Heavy Metal

Cost of Blanket Fabrication To be calculated

for each driver/
fuel cycle combination.

Cost of Reprocessing Spent Fertile Fuel $160/kg Heavy Metal

Cost of Shipping Spent Fertile Fuel $25/kg Heavy Metal

Plutonium Value (i.e., Pu Credit) @ $75/kg $33/gram Fissile
Separative lork

Cost of Spent Fertile Fuel Disposal $95/kg Heavy Metal

Cost of Waste Disposal (w/o Fissile Material) $20/kg Heavy Metal

Fusion Fuel Cycle Unit Costs

Cost of Deuterium $60/kg
Cost of Tritium $1,200,000/kg
Cost of Lithium $200/kg

IX-2
TABLE IX-A-1. (Continued)

Fusion Fuel Cycle Unit Costs (Continued)

Cost of Blanket Fabrication To be calculated
for each driver/
fuel cycle combination.

Cost of Reprocessing Lithium $100/kg

Cost of Shipping Lithium $25/kg

Accompanying Fission Reactors (LWR Complex)

Cost of Fission Reactor $650/ kWe

Cost of rabricating Mixed-Oxide Fuel $245/kg Heavy Metal
Cost of Reprocessing Spent Fertile Fuel $160/kg

Cost of Shipping Spent Fertile Fuel $25/kg

Operating and Maintenance Costs/Yr/MWe $5000

IX-3
TABLE IX-B-1. Capital Investment Cost Summary ($106)(a)(b)(c)

Driver/Blanket Pu Producing U-Pu Catalyst Refresh

Laser 2037 2775 1641
Mirror 2570 2991 2496
Theta-Pinch 2567 3797 2373
Tokamak 2074 2610 1990

(a) June 1978 price levels.

(b) Interest during construction and escalation during construction costs
not included.

(c) Hybrid reactor costs only.

TABLE IX-B-2. Capital Investment Cost Summary(3)(d)(€)(s)

Driver/Blanket Pu Producing U-Pu Catalyst Refresh

Laser 2167/kwe(bzc) 1770/kWe 1977/kWe
(617/kWth) (557/kWth) (544/kWth)
Mirror 18489/ kWe 5498/ kWe 35154/ kWe
(997/kWth) (830/kWth) (1038/kuWth)
Theta-Pinch 57044 /kWe 2438/ kWe Net Electricity
(531/kWth) (463/kWth) Consumer
(546/kWth)
Tokamak 2074 /kWe 1422 /kWe 2333/kWe
(501/kWth) (396/kWth) (536/kW th)

(a) June 1978 price levels.

(b) Net electrical output.

(c) Gross thermal output.

(d) Interest during construction and escalation during construction costs
not included.

(e) Hybrid reactor costs only.

IX-4
1) Switchyard and Transmission Facility

2) Escalation During Construction (computed in levelized energy cost

3) Escalation Prior to Construction calculations)

4) Decommissioning

5) Research and Development

6) Working Capital

7) Interest During Construction (computed in levelized energy cost
calculations)

Blanket costs are also excluded from the capital investment cost estimates
(included in the fuel cycle cost estimates). June 1978 price levels are
assumed in all estimates.

C. BLANKET COSTS

Blanket costs consist of the costs of (1) burchasing the structural
material and cladding components used in the blanket assemblies and 2)
fabricating the assemblies. The costs of blanket fuel materials are not
included as blanket costs. Structural material and cladding costs (i.e.,
material costs) consist of material purchase costs and the costs of material
losses during fabrication — in this study assumed to be 5% of the material
requirements. Fabrication costs include labor, assembly expense, and the
overhead costs on the plant and equipment used in the manufacture of the blan-
kets. A1l blanket costs are accounted for as fuel cycle expenses.

Material requirements for blanket assemblies are based on driver
geometries and blanket configurations. The middle regions of the blankets
containing fuel material, stainless steel, and Tithium dioxide were used
as the basis for the blanket cost calculations.

D. ANNUAL OPERATING AND MAINTENANCE COSTS (Hybrid Reactor)

Annual operating and maintenance costs are the routine day to
day expenses required to operate the reactor system. These expenses include
the costs of operating staff salaries, supplies, maintenance materials, and
process chemicals. Also included are the costs of routine maintenance and
replacement of major reactor components such as blanket assembly modules.
Estimated fusion-fission reactor annual operating and maintenance costs for
each of the reactor driver/blanket combinations identified in this study

are given in Table IX-D-1.
IX-5
TABLE IX-D-1. Annual Operating and Maintenance Cost Summary ($]06)(a)(b)
Pu Recycle/

Driver/Blanket Once-Through U-Pu Catalyst Refresh Cycle
Laser 41 56 33
Mirror 51 60 50
Theta-Pinch 51 76 A8
Tokamak 41 52 40

(a) June 1978 price levels.
(b) Hybrid reactor costs only.

E. FUEL CYCLE COSTS (Hybrid/Fission Reactor System)

Fuel cycle costs are the costs of operating the fuel material supply/
discharge cycle servicing the fusion-fission reactor system. For some
systems, fuel cycles are relatively simple, involving only fuel preparation
activities before fusion-fission reactor charging and spent fuel disposal
activities after fusion-fission reactor discharging. For other systems,
fuel cycles are more complex, some involving the coupling of fusion-fission
reactors into systems with conventional fission reactors — the fusion-
fission reactor producing the fissile fuel, the conventional fission reactors
consuming the fissile fuel. Regardiess of the complexity of the fuel cycle,
all system costs incurred for fuel material purchase, preparation, processing,
storage, transportation, and disposal are considered fuel cycle costs.

Estimated fusion-fission reactor fuel cycle costs for each of the
reactor driver/fuel cycle combinations identified in this study are given
in Table IX-E-1. Detailed fuel cycle cost estimates are given in Appendix B.
Costs are reported as levelized fuel cycle costs per unit of electricity
generated (see Section IX-F for description), Parameters and unit cost
assumptions used in calculations are given in Table IX-A-1.

XI-6
TABLE IX-E-1. Fuel Cycle Cost Summary (Mills/kup)(3)(b)

Priver/Fuel Cycle Once-Through Pu Recycle U-Pu Catalyst Refresh

Laser 8.6 3.8 3.0 -
Mirror 36.0 4.3 3.2 -
Theta-Pinch 1067.0 20.7 5.1 -
Tokamak 5.4 3.0 2.5 2.0

(a) June 1978 price levels.
(b) Complete hybrid/LWR system costs.

F. LEVELIZED ENERGY COSTS

Levelized energy cost or unit electricity cost is the average cost per
unit of generated electricity over the reactors operating lifetime (i.e.,
the average price that must be charged per unit of electricity generated
to recover all costs of constructing and operating the reactor system.
Capital investment costs, operating and maintenance costs and fuel cycle
costs are all expenses incurred in constructing and operating the reactor
system, and are therefore, used as input for levelized energy cost calcu-
lations. Estimated fusion-fission reactor system levelized energy costs
for each reactor driver/fuel cycle combination identified in this study
are given in Table IX-F-1. Detailed estimates of levelized energy costs
are given in Appendix B.

Levelized energy costs were estimated using the general economic condi-
tion input parameters and utility description data input parameters listed
in Table IX-A-1. Input parameters assume a real or deflated dollar analysis
(i.e., input parameters reflect values that would be found if there was no
inflation). A real cost of capital of 4%/year and no cost escalation were
assumed.

XI-7
Levelized energy cost estimates may vary considerably for similar sys-
tems due to differences in the input parameters and the estimating method-
ology used. This study used a discounted cash flow/levelized energy cost
estimating methodology.

TABLE IX-F-1. Levelized Energy Cost Summary (Hills/kun)(@)(b)(c)

Driver/Fuel Cycle Once-Through Pu Recycle U-Pu Catalyst Refresh

Laser 52.0 20.4 17.4 -
Mirror 405.1 31.2 21.6 -
Theta-Pinch 2205.8 37.5 19.2 -
Tokamak | 46.81 18.25 15.76 18.57

(a) June 1978 price levels.

(b) Complete hybrid/LWR system costs.

(c) Levelized Energy cost for a plutonium recycle LWR system is
15.2 mills/kWh.

The relationship between the annual unit cost of generating electricity
and the levelized energy cost is shown graphically in Figure IX-{-1. Annual
capital investment costs are fixed by the initial financing and are constant
over the systems operating lifetime. Operating and maintenance costs and
fuel cycle costs typically increase over time as affected by inflation and
real fuel price increases. As a result, the annual cost of generating
electricity increases over time. The levelized cost or levelized energy
cost is simply a present valued average measure of the increasing total
annual costs.

G. FISSILE FUEL VALUE (i.e., Breakeven Value)

A second criteria for judging the economic attractiveness of a particu-
lar fusion-fission reactor system is the value of the selling price of the

1X-8
B TOTAL
~ ANNUAL COST

[ LEVELIZED fOST / /

FUEL (IF THERMAL)

credes
SO XN KR
I I X, e,
QSRR IR

'a"a'e 46 5. 8.0.9.0,

COST OF ELECTRICITY (milts/kWh}

TIME (years)

FIGURE IX-F-1. Annual Cost of Electricity and Levelized Energy Cost

fissile fuel produced. This fissile fuel value or breakeven value is
defined as the price at which reactor breed fissile fuel could be sold
assuming the producing reactor is operating competitively, For reactor
systems with relatively small construction and operating costs, revenues
from fissile fuel sales need not be large to allow the reactor system to
operate competitively. Under these conditions, the fissile fuel value
would be Tow. For reactor systems with greater construction and opera-
ting costs, revenues from fissile fuel sales must be greater (to offset
increased construction and operating costs), resulting in higher fis-
sile fuel values. In this study, a fusion-fission reactor system (pro-
ducing fissile fuel) is assumed to be operating competitively if its
levelized energy cost is equivalent to the levelized energy cost of a
conventional LWR plutonium recycle system.

Fissile fuel breakeven values for both a fissile plutonium production/
sell fuel cycle and a fissile uranium production/sell fuel cycle for each
of the reactor drivers identified in this study are given in Table IX-G-1.

IX-9
TABLE IX-G-1. Fissiie Fuel Breakeven Values ($/gram Fissiie)

Fissile Fuel Va]ue(a)

Driver (Fissile Plutonium) (Fissile Uranium)
Laser 205 140
Mirror 525 299
Theta-Pinch 310 200
Tokamak 130 75

(a) June 1978 price levels.

H. MARKET PENETRATION

Fusion-fission reactor systems will not commercially deploy until
the present value benefits of a sustained commercial fusion-fission economy
become positive. Projections of the extent to which the fusion-fission
reactor systems identified in this study would deploy as commercial power
generating systems and the resulting economic benefits accruing to society
due to this deployment are described.

The benefits resulting from deployment are best measured as present
value benefits or present value energy generation cost savings resulting
from displacement of expensive alternative energy sources by cheaper fusion-
fission reactor systems. For this reason, the benefits resulting from
fusion-fission reactor system deployment are sensitive to the generating
costs of alternative energy sources. Two different alternative energy
source situations or sceanrios are examined in this study. These scenarijos
are described in Table IX-H-1.

TABLE IX-H-1. Energy Supply Scenarios

Scenario 1 Scenario 2
LMFBR Availability None 1993
CTR Availability 2010 2010
Fusion-Fission Availability 2000 2000
Electricity Demand Moderate/High Moderate/High

IX-10
Examination of the estimated costs of the fusion-fission reactor sys-
tems identified in this study revealed that none of the systems would be
economically competitive energy sources (i.e., all reactor systems were
estimated to be more costly to construct and operate than alternative
electricity generating sources). Therefore, this assessment of market
penetration potential is aimed at identifying the reduction in estimated
system costs that would have to occur before fusion-fission systems could
be projected to be competitive energy generation sources.

"Capitalized costs" are used as the aggregate measure of system con-
struction and operating cost. Capitalized costs are comprised of (1) the ini-
tial capital investment cost (i.e., plant cost), (2) the present value of all
fuel cycle cost streams (except purchase costs and sales revenues of nuclear
materials) over the systems operating lifetime, and (3) the present value of
all interim capital replacement cost streams over the system's operating life-
time. This cost measure does not include plant operating and maintenance
costs, costs or credits for electricity use or generation, nuclear fuel pur-
chases or credits, taxes, and insurance costs. Capitalized costs and other
measures of system performance are given in Table IX-H-2.

TABLE IX-H-2. Market Penetration Assessment(a) -
Economic and Performance Parameters

Estimated (b) Reactor Net

Capitalized Fissile Fuel Thermal Electric

Cost Production Power Output

Driver (106%) (kg/yr) (MWth) (MWe)
Laser 3190 1323 3300 940
Mirror 3680 807 2578 139
Theta-Pinch 6700 2592 4835 45
Tokamak 2696 1950 4144 1000

(a) Plutonium producing blanket assumed.
(b) June 1978 price levels,

IX-11
The extent to which the identified fusion-fission reactor systems can
be expected to deploy are given in Table IX-H-3 for the Scenario 1 energy
supply situation. Deployment projections assuming a Scenario 2 energy
supply situation are given in Table IX-H-4. Fusion-fission system deploy-
ment is stated both in terms of the number of fusion-fission power reactor
plants operating in the year 2030 and in terms of the present value bene-
fits through year 2040 of deployment, Reductions in estimated capitalized
costs required to make system projections look economically competitive
and deployable are also given.

I. NONPROLIFERATION IMPACT

When assessing the economics of proliferation resistant fusion-fission
systems, one characteristic quickly becomes evident. The additional costs
of utilizing nonproliferation mechanisms in fusion-fission systems, given
that the systems are developed with adequate planning and integration,
are not great relative to the total costs of constructing and operating
the systems. Preliminary estimates have indicated that when properly
implemented, proliferation resistant fusion-fission power generating systems
would yield power costs only 8% greater than power costs of fusion-fission
systems not specifically planned to be proliferation resistant. Five
mechanisms have been identified as candidates for making fusion-fission
systems more proliferation resistant. These mecharisms and their expected -
costs of implementation are discussed below.

1. Nuclear Center

The concept of an institutionalized nuclear center is very attractive.
Reduced nuclear material shipping distances resulting in a lessened
opportunity for nuclear material diversion is the primary advantage of such
centers. Increased system costs would be primarily due to increased
transmission distances. Given that centers are properly situated, decreased
costs could result from lessened licensing problems, lessened construction
and operating worker impacts, and sharing of common facilities. In
addition, cost decreases could result from shortened nuclear material
shipping distances and use of an integrated security system for all

IX-12
TABLE IX-H-3. Market Penetration Assessment/Scenario 1

Assumptions: 1) Moderate-High Electricity Demand

)
2) No LMFBR Availability
3)

)

4) Year 2010 Pure Fusion Availability

Number of 2500 MWth
(a)(c) Fusion-Fission Power
Capitalized Cost ¢} Reactors Operating

Year 2000 Fusion-Fission Availability

Driver ($/kuth) in Year 2030
Laser 965 (Estimated Cost) -

540 0

405 650
Mirror 1425 (Estimated Cost) -

210 0

145 900

Theta-Pinch 1385 (Estimated Cost) -

205 0
125 750
Tokamak 650 (Estimated Cost) _ -
540 0
430 600

(a) June 1978 Price Levels.
(b) 8.8%/yr Discount Rate,
(c) Plutonium Recycle Fuel Cycle

IX-13

PV Benefits(b)
to
Year 2040

Negative
0

10 Billion

Negative
0
10 Billion

Negative
0

10 Billion

Negative
0
10 Billion
TABLE IX-H-4.

Assumptions:

Driver

Laser

Mirror

Theta-Pinch

Tokamak

1) Moderate-High Electricity Demand

2) Year 1993 LMFBR Availability

3) Year 2000 Fusion-Fission Availability
4) Year 2010 Pure Fusion Availability

Number of 2500 MWth
(a) (c) Fusion-Fission Power
Capitalized Cost‘®/\C

Market Penetration Assessment/Scenario 2

PV Benefits'P’

(a) June 1978 Price Levels
(b) 8.8%/yr Discount Rate
(c) Plutonium Recycle Fuel Cycle

IX-14

Reactors Operating to
($/kWth) in Year 2030 Year 2040
965 {Estimated Cost) - Negative
440 0 0
365 550 10 Billion
1425 (Estimated Cost) - Negative
210 0 0
145 750 . 10 Billion
1385 (Estimated Cost - Negative
135 0 0
45 650 10 Billion
650 (Estimated Cost) - Negative
460 | 0 0
360 500 10 Billion
facilities. It is highly conceivable that the generating costs of a
proliferation resistant fusion-fission system located within a nuclear
center would be less than the costs of a decentralized fusion-fission
system with much greater potential for proliferation.

2. "Throw Away" Fuel Cycle

A second method for alleviating fissile material proliferation is to
"throw away" or dispose of the fusion-fission reactor spent fuel blanket
containing the fissile materials. However, such disposal would penalize
the fusion-fission systems as their primary function lies as fissile fuel
breeders. The specific economic penalty of utilizing a throw-away fuel
cycle can be approximated from results obtained in this study. Once-
through "throw away" fuel cycle systems are projected to operate at levelized
energy costs of 40 mills/kWh greater than reprocessing fuel cycle systems
(see Section F). Given an average demand for nuclear center generated
electricity between the years 2000 and 2030 of 1,000 GWe (2.6 x 10]4 kWh
cumulative), the economic cost of utilizing the proliferation resistant
"throw away" cycle between these years is in excess of 10 trillion doilars.

3. Co-processing

Co-processing is a third mechanism for reducing fusion-fission system
proliferation potential. Fusion-fission system cost reductions with co-
processing would result from lessened spent fuel reprocessing requirements.
System cost increases with co-processing would result from increased volumes
of radioactive fuel materials requiring remote handling, increased
transportation costs (due to additional fuel material volumes), and increased
re-enrichment and refabrication costs. In this study, the costs of
reprocessing, transportation, and fuel fabrication in a plutonium recycle
system are estimated to make up only 8% of the system's power cost. Given
that fusion-fission system fuel cycle operations are planned and integrated,
the additional costs of including co-processing in fuel cycles should increase
power costs by less than this 8%.

IX-15
4. Refresh Blanket

The refresh fuel cycle/blanket concept is a fourth mechanism for
retarding fusion-fission system proliferation potential. In this fuel cycle
concept, the fuel blankets are laden with fission and activation products
making them hiahly radioactive and providing themselves proliferation
resistance. Results obtained in this study indicate that the additional
costs of utilizing such a nonproliferation device are negligible (see
Section F}).

5. Denaturing

233U or 235U, using 238U as a dilutant provides

a fifth mechanism for obtaining proliferation resistant fuel cycles. Like

Denaturing of fissile
co-processing, this mechanism affects only the reprocessing, transportation,

enrichment, and refabrication stages of fuel cycles resulting in a maximum
impact on system levelized energy costs or power costs of E%.

IX-16
J. SECTION IX REFERENCES

1. S. C. Schulte, 7. L. Wilke and J. R. Young, Fusion Reactor Design Studies -

Standard Accounts for Cost Estimates, PNL-2648, Battelle Pacific Northwest
Laboratories, Richland, WA 99352, May 1978.

2. Guide for Economic Evaluation of Nuclear Reactor Plant Designs, NUS-531,
United States Atomic Enerqgy Commission, Division of Technical
Information, Washington, D.C., January 1969.

3. Laser Fusion Hybrid Reactor Systems Study. UCRL-13720, Lawrence
Livermore Laboratory, Livermore, CA, July 1976.

4, Laser Fusion-Fission Reactor Systems Study, UCRL-13796, Lawrence
Livermore Laboratory, Livermore, CA, July 1977.

5. T. H. Baltzer, et al., Conceptual Design of a Mirror Reactor for a Fusion
Engineering Research Facility, UCRL-51617, Lawrence Livermore Laboratory,
Livermore, CA, August 1974.

6. B. W. Moir, Conceptual Design of a Mirror Hybrid Fusion-Fission Reactor,
UCRL-51797, Lawrence Livermore Laboratory, Livermore, CA, June 1975,

7. D. J. Bender and G. A. Carlson, System Model for Analysis of the Mirror
Fusion-Fission Reactor, UCRL-52293, Lawrence Livermore Laboratory,
Livermore, CA, October 1977.

8. K. R. Schultz, et al., Conceptual Design of the Blanket and Power
Conversion System for a Mirror Hybrid Fusion-Fission Reactor, GA-A14021,
General Atomic Company, San Diego, CA 92138, July 1976.

9. Krakowski, et al., Engineering and Physics Considerations for a Linear
Theta-Pinch Hybrid Reactor (LTPHR), Los Alamos Scientific Laboratory, Los
Alamos, NM, June 1976.

10. A Feasibility Study of a Linear Laser Heated Solenoid Fusion Reactor,
EPRI-ER-171, Electric Power Research Institute, Palo Alto, CA, February
1976.

11. J. K. Ostic and J. Sheppard, Synfuel and Fissile Production from a Linear
Theta-Pinch Fission/Fusion Hybrid Reactor, University of Washington,
Seattle, WA, June 1977.

12. B. Badger, et al., UMAK-III, A Noncircular Tokamak Power Reactor Design,
UWFDM-150, University of Wisconsin, Madison, WI, July 1976.

IX-17
13.

14.

15,

16.

17.

18.

B. Badger, et al., TETR, A Tokamak Engineering Test Reactor to Qualify
Materials and Blanket Components for Early DT Fusion Power Reactors,

UWFDM-191, University of Wisconsin, Madison, WI, June 1977, Rev. December
1977.

Experimental Fusion Power Reactor Conceptual Design Study. EPRI ER-289,

Electric Power Research Institute, Palo Alto, CA, December 1976.

The Handy-Whitman Index of Public Utility Construction Costs. Whitman,

Requardt and Associates, Baltimore, MD. 1977.

D. E. Deonigi and R. L. Engel, Performance Targets for Fusion-Fission
(Hybrid) Reactors. BNWL-2139, Pacific Northwest Laboratory, Richland, WA

99352, January 1977.

V. L. Teofilo, et al., "A Tokamak Hybrid Blanket Design," Proceedings of
the Seventh Symposium on Engineering Problems of Fusion Research, IEEE

p. 6, No. 77CH1267-4-NPS, p. 1624, Knoxville, TN, October 1977.

V. L. Teofilo, "Tokamak Demonstration Hybrid Reactor," in Proceedings of
Second Fusion-Fission Energy Systems Review Meeting, Washington, D.C.,

p. 513, November 1977.

IX-18
X. LICENSING AND SAFETY

A wide variety of hybrid concepts is possible, as seen in this report in
the discussion of alternate fusion drivers, reactor coolants, fuel forms, and
fuel cycles. So before a detailed discussion of the specific licensing and
safety issues associated with each of the four drivers considered is given, a
generic discussion of the problems facing the hybrid concept in general is in
order.

A. GENERIC DISCUSSION OF THE HYBRID CONCEPT

The fusion-fission hybrid reactor concept is based on the energy gain
realized when neutrons produced by a thermonuclear deuterium-tritijum plasma
interact with a surrounding blanket containing fissile or fissionable material.
Again, a large array of concepts is possible among the magnetic and inertial
confinement fusion driver concepts available, as well as the choice of fission
fuel cycle, heat removal cycle, etc. Typically, the fusion reaction is con-
fined to the interior of a large vacuum vessel (torus or cylinder, etc.),
which then dictates the general blanket geometry.

The fission zone is usually quite thin (~ 1 m or less); however, because
of the size of the "shell" structure it is typically divided into quadrants,
often with independent coolant loops, and further divided into modules or
assemblies. The placement of energy systems required to initiate the fusion
reaction (superconducting magnets, injector or beam lines, cryopanels, etc.)
and the routing of the cooling system adds further complexity to the structure,
often resulting in complex shapes with access problems for fabrication,
refueling and maintenance.

In addition to the above, the blanket region must breed sufficient tritium
for operation of the fusion driver. Lithium or lithium compounds must then
be included, sometimes in the form of a liquid metal, where it can also operate
as a reactor coolant.

X-1
Major design features include the lack of any reactivity controls. The
fission blanket itself is designed to be subcritical over the fuel Tifetime,
and any large excursions in fusion neutron production are considered highly
unlikely, Power densities are usually below those found in pure fission
reactors, and the segmented blanket design tends to isolate coolant flow

disturbances.

X-2
B. GENERIC SAFETY AND LICENSING ISSUES

1. Radiation Exposure

Many of the safety and licensing aspects of hybrid plants will focus on
the presence of radioactive materials, which will include tritium, activation
products, fission products and actinides. These are found to varying degrees
in modern fission reactors; however, for safety analysis and licensing the
specific radionuclide inventories as well as their chemical form and location
in hybrid systems must be identified. In addition, it must be demonstrated
that the hybrid blanket modules can operate safely in close conjunction with
high energy fusion systems. Of concern are unique initiating events leading to
loss of containment as well as the identification of routine occupational
exposure during plant operation, refueling and maintenance. A short discussion
of the various radioactive materials present will now be given.

a. Tritium

Due to the lower energy balance constraints placed on the fusion driver,
a hybrid system may be the first commercial application of the D-T fuel cycle.
To meet daily requirements, an extraction and separation process will probably
require tritium to be present in kilogram quantities in the blanket.

Tritium (T) is a radioactive isotope of hydrogen which decays by emitting
a soft beta particle (E = 5.7 keV, Eax = 18 keV) and no gamma ray, and is
therefore a significant radiological hazard only if ingested. Since T2 is
virtually insoluble in human tissue (about 98% of T2 inhaled is immediately
exhaled), it is relatively innocuous. Tritiated water (T20, HTO or DTO),
however, is a much greater hazard. The maximum permissible concentration (MPC)
value for tritiated water in air is 0.2 uCi/m3 (uncontrolled area), 1/200 of
the comparable value for T2.

Research and development is therefore required for tritium monitors capable
of discriminating between molecular tritium and tritiated water and of accurate
real-time measurement of tritium concentrations on the order of 0.1 uCi/mB.
Without this development all tritium detected in the facility atmosphere must
be assumed to be tritiated water. Such an assumption will decrease design and
operational flexibility and increase costs.

X-3
In-plant tritium releases during normal operation would primarily result
from leaks, particularly around valves, greatly exceeding contributions from
permeation. One cause of leaks is the damage to elastomeric seals resulting
from tritium exposure. The identification of tritium-resistant materials
should proceed. For maintenance purposes, every tritium handling component
should be designed so it can be purged. Components contaminated by tritium
alone, however, do not require remote maintenance: a combination of glove boxes,
plastic tents and bubble suits with independent air supply will be adequate
for maintenance operations.

Design parameters for emergency cleanup systems will depend on the accident
scenarios identified and the form of tritium released as discussed above. The
conversion rate of T2 to tritiated water (mostly HTO) will be a strong function
of environment in the reactor hall (e.g., surface conditions, temperature,
humidity, etc.). The identification of design basis accidents will be dis-
cussed under the section on accidents.

The safety and licensing aspects of large-scale tritium use are being
investigated for the magnetic fusion program; however, the specific tasks and
schedules for this research may have to be reevaluated for early applications
in hybrid systems.

b. Blanket and Structure Activation

D-T fusion drivers, as copious sources of neutrons, will activate struc-
tural, blanket and shielding materials with profound effects on overall machine
design, operational planning, and costs. In particular, maintenance operations
on components within or proximal to the fusion device will be affected. Most
near term concepts project the use of stainless steel, which will surely make
a substantial remote maintenance capability necessary. The fusion structures
typically require replacement after several years of operation, and so the
vacuum vessel must be designed with remote cutting and disassembly in mind.

The mass transport of activated structural materials in the coolant system
(corrosion) and in the vacuum system (evaporation, sputtering, blistering, etc.)
must also be considered. In fission reactors, unforeseen radioactive crud
buildup in areas requiring maintenance is often the major source of occupational

X-4
exposure. Test loops to identify problems in 1iquid Tithium cooled systems
are now underway. Mass transport in vacuum systems may require some operational
experience to pinpoint problem areas.

¢. Fission Products and Actinides

The presence of fission products and actinides in the hybrid blanket have
three aspects which require unique investigation. First, the hard fusion
neutron spectrum is likely to generate different radionuclide concentrations
for the many fuel cycles and fuel types (carbides, oxides, metals, salts, etc.)
under consideration for hybrids. Licensing considerations then require that
the research include the following for all fuel combinations:

e establish nuclear data files for fusion spectrum,
e determine radionuclide inventories at exposure,
e determine decay heat curves.

The nuclear data is being formulated today; however, all hybrid neutronics to
date have relied on fission reactor spectra and light water decay heat curves.
Although this type of analysis is an acceptable approximation for today's
design studies, it could not serve as the basis for component design of decay
heat removal systems for actual systems subject to regulatory review.

The second area requiring work deals with the mechanical performance of the
fuel in a hybrid application. Many fusion drivers operate in a cyclic or
rapidly pulsed mode, resulting in thermal and radiation conditions far different
from those found in fission reactors which operate in a relatively steady state
mode. Commercial applications will require that the fuel be fully qualified
in the hybrid environment during startup, operation and shutdown of the reactor.
As with fission reactors, the hybrid would then be licensed to operate within
a specific performance envelope defined by the fusion driver characteristics
and fuel response. Fuel failure rates are also required to identify circulating
inventories in cooling systems for accident analysis.

Finally, the shape of the fission blanket itself will introduce problems
for refueling and maintenance. Many hybrid designs use large, irreqularly
shaped fuel modules which are welded into the reactor structure and cooling

X-5
system. Refueling then requires remote cutting and welding and the transport
of large assemblies. Safety analysis will be required to provide input into
reactor design to minimize exposure during these operations.

2. Accidents

It will be in the area of accident analysis more than anywhere else that
the formulation of regulatory codes and design standards and materials quali-
fication will impact the licensing of hybrid systems. It is recognized that
this will be an iterative process, with initial scoping studies forming the
basis for early design requirements. These will then eventually form the basis
for the regulatory licensing functions which must provide the following:

® design basis accidents,
e analysis codes and assumptions,
e design standards and criteria.

Regulatory review will interact between preliminary and final safety systems
design and, of course, update standards on the basis of operational experience.

Much of the unique accident analysis and safety design work required for
licensing hybrid systems will deal with the containment of the radioactive
materials just discussed. Of particular concern are initiating events leading
to loss of coolant in the fission blanket, possibly followed by fuel melting
and loss of containment, or accidents affecting the sub-critical nature of the
blanket. It must be demonstrated that the fission blanket can operate safely
in close conjunction with any high energy fusion systems.

a. LOCA/LOFA Accidents

At this stage of hybrid development it is difficult to identify initiating
events in the various conceptual blanket designs which could lead to local flow
reduction or blockage events, or more serious accidents involving larger
portions of the cooling system. Analyzing the blanket response to a postulated
event has the same problem due to the complex structural geometry. It is
thought that the modular design of most hybrids with independent cooling loops
serving a quadranted blanket will tend to isolate disturbances making it
unlikely that fuel melting will propagate., However, this geometry distributes
the fission blanket over a large region making it difficult to provide guard

X-6
vessels around all structures for containment if melt-through does occur.
Because of this Tocalized melting may still result in widespread contamination
of other reactor systems. Also, because no one portion of the blanket can be
isolated from the fusion driver during operation, a large instrumentation
system will be required to spot isolated cooling problems which would require
a power reduction in the entire blanket.

The response of a hybrid blanket to a loss of flow or coolant type accident
will depend directly on the type of coolant, the operating power density, the
speed with which reactor shutdown can occur, and the decay heat levels which
were discussed earlier. Initial hybrid designs had very low power densities
(10-20 w/cm3) making decay heat cooling by natural convection with 1liquid metal
systems possible if designed with a functioning heat sink. For more recent
designs, average blanket power densities have increased significantly with most
relying on helium coolant. The power densities are well within modern HTGR
and GCFR technology; however, forced circulation must be maintained with the
gas cooled hybrid design to prevent fuel melting from decay heat. Another
problem with the gas cooled designs is that the system has very little thermal
inertia, making rapid shutdown of the fusion driver (and the initiation of
auxiliary cooling if possible) imperative in a loss of flow accident. Rapid
shutdown is easily achieved with inertial confinement and some pulsed magnetic
fusion drivers; however, tokamaks or mirrors may require a significant cooldown
period to quench the fusion reaction and avoid damage caused by a plasma dump.
A safety system consisting of an emergency injection of impurities or an over-
fi1l of hydrogen may be required to improve the shutdown response for large
plasma devices.

b. Criticality

One of the major safety objectives of the hybrid design is to insure that
the fission blanket remains subcritical under all conditions. Although the
blankets are designed to be subcritical over the entire fuel lifetime, various
mechanisms are available for reactivity insertion. If criticality could be
achieved, large power excursions and energetic disruptions leading to large
scale release of radionuclides can be envisioned. This possibility must and
can be eliminated.
Changes in blanket geometry caused by gross physical displacement (e.g.,
collapse of structures),or by fuel melting in LOFA/LOCA accidents, or by the
introduction of steam or water in blanket voids are considered to be the most
serious ways of changing blanket reactivity. The hybrid blanket differs from
pure fission reactors in that it is structured around the fusion driver vacuum
vessel and is far from being in its most compact geometry. Also, most hybrid
fuel cycles are designed to be breeders, with fissile fuel content increasing
with exposure. The blanket response to various reactivity insertion accidents
then becomes more serious with time.

The criticality calculations done for hybrids today are highly conservative
in that they typically assume total collapse of the fission blanket. Plotting
the Kéff resulting from this "accident" as a function of blanket exposure
(fissile fuel content) then defines the useful blanket lifetime to keep Keff < 1.
Criticality calculations used to date for fuel meltdown accidents follow the
same pattern, where reconfiguration is assumed to be as a sphere which is the
most reactive geometry.

Steam ingress accidents for gas cooled designs have the potential for
neutron thermalization in a blanket designed for a fast spectrum, possibly
leading to criticality with a sufficient fissile buildup. However, this requires
the accident to progress from steam leakage in the blanket to failure of the
blanket with steam and water filling the vacuum vessel, followed by over-
pressurization and expansion of the blanket. With low burnup blankets the
volume of water required to achieve criticality is estimated to greatly exceed
steam generator inventories.

Greatly increased neutron output from the fusion reaction is not seriously
considered as a source for reactivity input. Due to the difficulty in initiating
the reaction, below par performance of the fusion driver will more likely be
the case.

Motion in the fuel assemblies caused by thermal bowing or flow induced
vibrations will not be unique to hybrid designs.

X-8
The conservative criticality calculations done to date then indicate that
hybrid designs are possible that eliminate the chance of criticality, and
that this can be considered an inherent safety feature. However, it again remains
to establish more realistic design accidents and analyses to allow for the
optimization of blanket performance (Keff) while still retaining this feature.

c. Vacuum Vessel Safety

The presence of the large vacuum vessel in the center of the fission blanket
has been mentioned several times. This structure is often used to provide

support for the blanket as well as containing the fusion reactions. As such
its failure is capable of affecting the integrity of cooling systems and fuel
geometry, as discussed earlier for LOFA/LOCA accidents and criticality.
Missile generation upon failure could also affect the cooling system and
possibly the magnet systems for magnetic fusion devices.

Licensing considerations would then be directed towards appropriate
materials qualification. Engineering for the necessary structural support is
not foreseen as a serijous problem {(although designing for access and maintenance
may be); however, the lead time required to qualify new materials may be
substantial. For example, it now takes approximately eight years to qualify
a new material or alloy for the ASME boiler codes.

d. Hazardous Materials

Finally, materials which present occupational hazards or accident potential
are used in the fusion driver systems. A major concern is the explosive nature
of hydrogen which is used in all fusion drivers and its potential for releasing
tritium. Hydrogen contains a great deal of potential energy; it contains 60,000
Btu/1b vs 20,000 Btu/1b for gasoline and 17,000 Btu/1b for dynamite. There is
a 90% chance that hydrogen leaks will ignite spontaneously under certain con-
ditions. Hydrogen will auto-ignite at 585°C.

The various design solutions suggested are:

e Use of surge volumes and/or rupture discs.

e Double walled, inert atmosphere tritium transfer lines.

e Explosion-proof electric motors and coated wires in tritium facility
buildings.
. H2 detectors, 1.5% turnoff source and sprinkler initiators.
e Limit combustibles.
e High hazard volumes--Halon (CBF3) explosion suppressors.,

The advantages and disadvantages related to the use of an inert atmosphere will
have to be resolved.

Other safety and licensing issues impacting accident analysis or occupa-
tional safety are associated with the fusion driver; however, these tend to
be design dependent (1iquid 1ithium, magnets, laser 1light, etc.). Such issues
will be addressed in the following discussion of the Tokamak Hybrid Reactor if
applicable.
C. TOKAMAK HYBRID

1. Description of the Tokamak Hybrid Concept

The main fusion driver for the tokamak hybrid presented in this report is
based on the Tokamak Engineering Test Reactor (TETR) designed by the University
of Wisconsin. This pure fusion device has been modified by adding a helium
cooled first wall with a surrounding fission blanket. This particular design
has been designated the Tokamak Hybrid Reactor (THR).

In the THR the thermonuclear deuterium-tritium plasma is confined magneti-
cally in a toroidal vacuum chamber with a major radius of 5.4 m and a minor
radius of 2.4 m. A double-null poloidal divertor directs impurities to particle
collection plates with final vacuum pumping being done by cryo-sorption panels
located in the divertor region. Cryosorption panels are also used in the neutral
beam injection ports which introduce penetrations around the circumference of
the torus. The entire vacuum vessel and divertor regions are @ncompassed by
the toroidal field "D" magnets assumed to be superconducting niobium-tin in this
design. The inner edge of all "D" magnets is then attached to a center support
stanchion. |

Due to the lack of access between the vacuum vessel and this inner stanchion,
no fission assemblies are located here, usually just shielding to protect the
magnets. However, in this particular design flowing natural 1liquid 1ithium
has been added to supplement the tritium breeding.

The fission blanket is then restricted to slightly less than 180° of the
outer poloida] angle of the vacuum vessel. It is divided into segments around
the torus in "orange pée]" fashion, with approximately three segments per
toroidal field coil. The final design number of "D" coils and blanket segments
has not been established. Each segment is then further divided into the 11
modules.

In the PNL hybrid modification of the TETR, the original steam-cooled
stainless steel tubular first wall facing the plasma is replaced by a thin
stainless steel water-cooled double wall with a carbon liner. The blanket
mocdules plug into the helium delivery and collection ducts directly behind
the modules. Stainless steel cladding is specified for the fuel rods and L120
contained in the mocdules. Shielding is then placed outside of the blanket
assemblies where more field shaping coils are located.

The power conversion system for the THR has not been specified yet, but
is 1ikely to consist of four independent primary cooling loops, each with two
main steam driven helium circulators. An auxiliary cooling system for each
loop with electric driven circulators and its own independent heat exchangers
would provide backup or emergency decay heat removal. Both systems should
be capable of providing adequate decay heat removal independently in a
depressurization accident.

2. Safety and Licensing Issues for the THR

As mentioned in the generic discussion of safety and licensing issues for
hybrids, one of the initial tasks in safety analysis of the THR will be to
identify those unique operating characteristics or systems which may impact
accident analysis.

The fusion driver for the THR operates in a cyclic mode with plasma heating
lasting three seconds, followed by approximately 100 seconds of plasma burn
and a ten second cooldown and refueling cycle. With a driven tokamak, the
temperatures required for the fusion reaction are maintained by beam injection
and resistive heating; however, this design assumes an ignited plasma capable
of maintaining the fusion reaction by utilizing the 3.5 MeV alpha energy.

The impact of continued fusion energy production in loss of coolant type
accidents must then be addressed, along with various methods of rapidly
quenching the fusion plasma. Undoubtedly the best approach will be the injection
of impurities or cold hydrogen fuel to luwer the plasma temperature. An emergency
loss of confinement with a subsequent plasma dump to the first wall is another
possibility, but has the potential for causing significant damage. For
example, a highly localized dump has the potential for melting the first wall
and dumping high pressure steam into the toroidal vacuum chamber. Tempera-
tures for a bare stainless steel THR wall could exceed ~ 1000°C - in 0.2 seconds
with a dump of 1500 w/cmz. The carbon curtain gives added protection but
can be vaporized in ~ 10 seconds with a dump over 1000 w/cmz. (Accidental
disturbances in magnetic confinement and magnet failure have safety implications
themselves which will be addressed below.).

X-12
The operating power levels and decay heat curves for the various fuel
cycles proposed for the THR will help determine an appropriate fusion driver
response to disturbances in the cooling system. No decay heat curves have
been calculated yet for the THR due to a lack of proper neutronics data.

Even the more recent safety evaluation of a gas cooled mirror hybrid desiqn(z)
relied on decay heat standards for thermal reactors.(3) The calculations
required for the various fuel cycles in THR are:

e time to fuel damage with reactor at full power following LOFA.
o time to fuel damage following LOFA and shutdown from full power.
e time to fuel damage following LOFA 48 hours after shutdown (refueling).

The qualification of fuel pin failure rates in the cyclic THR power cycle will
also be required for licensing. As with all gas cooled reactors, the immediate
hazard associated with reactor coolant leakage will be the radiological exposure
due to coolant-borne tritium and fission products that leak out with the coolant.
Fission products leaking from defective fuel pins plate out on the internal
surface of the helium loop, with the potential for being lifted off and blown
into the containment building during depressurization. The circulating tritium
activity was expected to be comparable to the fission product activity; however,
due to the lower radiological toxicity of tritium it does not contribute
significantly to the hazard potential in this type of accident. The actual
circulating tritium inventory in the THR design will depend on the character-
istics of the Lizo canisters used in the blanket modules.

In the THR design an evaluation of tritium containment and cleanup require-
ments must be extended to the 1liquid 1ithium breeding region in the central
support region of the tokamak. So two different tritium process streams must
be evaluated along with collection in the torus vacuum system and in the
reactor coolant,

Again, as mentioned in the generic discussion of issues, some hybrid
concepts place high energy fusion systems in close proximity to the irreqularly
shaped blanket modules introducing the anticipated problem of identifying

X-13
realistic accidents and predicting the system response. Examples with the THR
are the liquid lithium region and the use of superconducting magnets.

3. Liquid Lithium Spills

Accidents affecting the integrity of the THR structure could cause the
liquid T1ithium to be released. The 1ithium would then be very hot and chemically
very reactive, and could cause damage to components that it contacts directly,
such as shielding, structural supports or magnet components. At high tempera-
tures it can ignite spontaneously in the air and would react vigorously with
water and concrete. Lithium fires can then cause further damage directly,
or lead to overpressurization and missile generation which may damage other
blanket components and containment. |

Experimental programs for sodium-concrete and sodium-steel-concrete inter-
actions, in support of LMFBR safety, are available to illustrate methods for
treating lithium spills. The likelihood of serious lithium spills can be
reduced by utilization of a number of safety features, such as maintaining an
inert atmosphere outside the lithium loops and providing double-walled piping.

A number of major research projects have been suggested for 1ithium safety
in the magnetic fusion program. These include the following areas:

¢ Jlithium-concrete reactions

¢ J]ithium-material reactions

e Tithium spill extinguishment

e J1ithium aerosol behavior

e 1lithium air cleaning ccncepts

e water/gas release from concrete

e hydrogen formation

e liner concepts

e use of sodium safety analysis codes

Many of these areas are, in fact, planned for investigation in the current
program at the Hanford Engineering Development Laboratory in Richland, Washington.

The point to be made with the THR design is that the use of liquid 1ithium
only tc supplement tritium breeding will still require a major safety and materials
qualification program for licensing. The exclusive use of L120 pins would
probably be more attractive for near-term reactor applications.

X-14
4. Magnet Safety

The superconducting toroidal field coils for the THR carry megajoule
energies in close proximity to the fission blanket. The major safety concerns
for magnet systems will then include:

e Jjoule heating within a magnet or conductor sufficient to vaporize
material.

¢ sudden helium vaporization from heating resulting in destructive
rupture of the helium coolant system.

e thermal stress ruotures of magnets.

e electric arcing with material vaporization and generation of high
temperature flying material.

e generation of eddy currents and stray electric fields.

For licensing purposes the above research would form the basis for identifying
accident initiators whether they originate in the magnet or external to the
magnet in other hybrid systems. It would also produce the realistic assumptions
and codes to be used in accident analysis, and the criteria for engineered
safety features in magnet design. A large superconducting development program
is underway for the magnetic fusion program, and on the basis of safety studies
at BNL various engineered safety features can be envisioned for reactor appli-
cations as shown in Table X-C-1.

Again, if the THR represents a near-term first application of fusion
driver technology, the timetable for the above work would have to be adjusted
accordingly.

5. Criticality

The Keff of specific THR blankets has not been calculated; however, the
performance of a Pu catalyst fuel cycle can be extrapolated from previous
designs. In Reference 4 the Keff of the blanket started at 0.9441 and
increased to 0.9582 after two years of operation. This would be quite high
for an initial hybrid application where design studies typically put
Keff = 0.5. To operate the fission blanket at the Keff would imply that the
THR be designed to some standard for reactivity insertion accidents beyond a
simple "total collapse and melting to a sphere" type of calculation. This
further implies that a significant amount of research into design basis

X-15
TABLE X-C-1. Engineered Safety Features for Fusion Magnets

Type of Engineered
Safety Feature Function

Detection Systems e Detect local hot spots in coil.
e Detect lead overheating and failure.
e Detect arcs in coil.
e Detect loss of coolant or flow.
e Detect excessive strain or movement.

Temperature Equilibration e Drive all conductors normal early
Systems in a quench

¢ Remove coolant rapidly.
Energy Removal Systems ¢ Dump coil energy in external resistance.

Energy Dispersion Systems e Prevent excessive local deposition of
coil energy.

Containment Systems e Prevent or minimize coil disruption
consequences if coil winding fails.

accidents and the assumptions for accident analysis must precede the THR design
to the point where the standards and criteria have been accepted as the basis
for regulatory review and licensing. Blanket reactivity for THR in the

Keff = 0.5 -~ 0.6 range would be more realistic for initial applications of
hybrid technology.

6. Magnetic Fields

Magnet safety must also address the issue of occupational exposure to
high magnetic fields. The magnetic fields resulting from operation of a
fusion driver may have strengths up to several hundred kilograms with pulse
durations from several msec to hours and duty cycles of up to 80%. Fusion

plant employees could then be subject to high magnetic fields throughout their
work period.

Numerous studies have been made to determine the biological effects to
humans of magnetic fields. These studies include cardiac function, respiratory
function, behavioral changes, food consumption and growth, fetal development,
brain electrical activity, pathologic changes in spleen, liver, adrenal and

X-16
bone marrow, metabolic rates, hematology (red blood cells and leukocytes),
antibody production, wound healing, tumor growth, cell culture (growth and
function), cell division, genetics, enzymes, neuromuscular function, and survival.
However, the results from these studies are ambiguous; for example, the results
for several experiments on cell culture growth are about equally divided
between no effect, increased growth, and decreased growth. Such results could
be due to the normal range in experimental results, failure to control or
measure important variables, or some unknown reason.

A series of closely controlled experiments has been recommended for the
magnetic fusion program to determine the effects of exposure to magnetic fields.
Typical biological effects that should be studied are:

e neurological and behavioral phenomena
e life span exposures
o effects on development

- teratologic studies

- reproductive performance

- postnatal performance after prenatal exposure
e sStudies of combined insult

- radiation

- drugs or dietary alterations

-~ smoking

- chemical carcinogens
¢ epidemiologic
avian

mechanistic

In addition, there is a need for development of a personnel dosimeter.

The results of the above studies should be used to reevaluate the standards
for exposure to magnetic fields. The U.S. and some foreign nations have
established standards; however, the U.S. standards are less stringent by several
orders of magnitude.

7. Cryogenics

Cryogenic systems find a number of applications in the THR fusion driver
in addition to the superconducting coils just discussed. Specificaliy, cryogenic

X-17
condensation or sorption panels are often specified for vacuum systems due to
their high pumping speeds at low pressures. They also find application in
tritium containment and separation by distillation. These systems may then
be subject to failure resulting in extreme temperature or pressure excursions
capable of damaging other components or injuring personnel.

As with the superconducting magnets, the cryogenic systems will then
require engineered safety features to both detect local heating or pressure
increases, and containment systems to minimize the effect of loss of cryogenic
fluids. Again, studies will be required to determine potential accident
initiators and resulting consequences.

8. Activation Products

As for activation products, the production of radioactive materials in
the stainless steel first wall was calculated for the original TETR design.
The activity peaks at v 0.5 Ci/watt of fusion power generated after several
years of operation. The use of helium is expected to minimize the problem of
corrosion and activation product transnort in the coolant system. The use of
a carbon liner on the vacuum first wall is also expected to reduce the erosion
and transport of stainless steel activation oroducts in the vacuum system.
However, the liner requires periodic replacement. At the high activation
levels expected, remote maintenance will be required. It is not expected that
any special radiation exposure standards will be required. However, it is
obvious that a great deal of analysis into the methods of remote fabrication
and disassembly will be required to demonstrate that compliance with radiation
standards can be achieved.
D. MIRRQR HYBRID

1. Description of the Mirror Hybrid Concept

The reactor description here is based on the Lawrence Livermore, General
Atomic designgn'This design is helium cooled, with the magnet coils, blanket
and primary heat-transfer loop all located within a pre-stressed concrete
reactor vessel (PCRV) of the type developed for gas cooled fission reactors.

The primary consideration for the PCRV is to provide a high level of confidence
that forced cooling to the blanket can be maintained in accident situations.
The PCRV also provides the main restraining forces for the magnet. Thermal
insulation must be provided between the concrete of the PCRV and the super-
conducting magnet, which operates at 4°K.

The PCRV terminates in a hollow spherical region located within the
Yin-Yang magret coils, with penetrations for beam injectors and particle
streaming for direct conversion. The helium ducts‘are laid in as an integral
part of the PCRV, and then terminate in the central spherical hollow area.

The helium delivery and return ducts then connect to a spherical manifold sys-
tem which is suspended directly from the inside wall of the PCRV. This structure
also forms the vacuuvm vessel for plasma containment and is water cooled.

The fission blanket then consists of small modules which bolt directly to the
manifold wall, using a double knife-edge (Varian-type) seal to prevent gas
leakage to the vacuum chamber.

For tritium breeding, the original design called for lithium deuteride
pins in Lockalloy 43 cladding. This has been replaced for this report by
1ithium oxide pins with stainless steel cladding.

One of the main goals of this study was to investigate the feasibility of
applying gas-cooled reactor technology to the mirror hybrid. Helium is then
used as the main reactor coolant, with the system consisting of four independent
primary loops and four independent auxiliary loops. The primary loops are used
for normal power operation and for shutdown or depressurized cooling, with the
auxiliary loops used for reactor decay-heat removal following normal or
emergency loss of the primary loops. The blanket is divided up into four
quandrants, with a total of eight primary helium circulators and eleven steam
generators. The auxiliary system consists of five circulators and five auxiliary
heat exchangers. The primary helium circulators are steam driven, where the |
auxiliary circulators are electric driven.
The design includes a large vacuum chamber below the reactor for direct
conversion of plasma streaming.

2. Safety and Licensing Issues for the Mirror Hybrid Reactor

a. LOFA/LOCA

As with the gas-cooled tokamak hybrid, one of the major safety con-
cerns of the mirror hybrid reactor (MHR) will be in assuring the integrity
of the cooling system under accident situations. Forced convection again
must be maintained to prevent fuel melting. The mirror hybrid reactor has
a major safety advantage in that it uses the prestressed concrete reactor
vessel (PCRV) technology developed for the gas-cooled fission reactors.

The entire primary heat transfer system, including the steam generators,
delivery and return lines and manifolds for the fission blanket, are either
incased in or attached to this structure. It is stated that no damage

or malfunction may be incurred in this system by internally generated

(flow induced) or external vibrations. This reinforced structure also
protects the cooling system from possible accident scenarios involving the
superconducting coils, which would otherwise surround the coolant mani-
folds to the blanket. It is unlikely that failure of coils leading to
missile generation would then affect the integrity of the cooling system.
Unique safety research for the MHR would then 1ikely be focused on local
coolant flow disturbances or blockage accidents in the fuel modules themselves.
Mechanisms that could lead to propagation of failure from fuel rod to fuel
rod identified for the MHR include:

® Relocation of debris in adjacent cooling channels and
on spacers.

e Melt-through failure of wall leading to coolant bypassing
and flow reduction to modules.

e Reactivity changes due to blanket material relocation
leading to power increases and acceleration of failure
development.

This makes the rapid detection of coolant flow disturbances imperative.
Instrumentation will then be required to monitor module coolant outlet

X-20
temperature and activity levels and flux monitoring for power levels. However,
the response time of the instrumentation must be capable of providing an
unambiguous signal before damage can propagate.

A conservative (assuming a diabatic heat-up) estimate of the LLL/GA mirror
hybrid puts the time available before fuel damage occurs in loss of cooling at
full power at 1.5 seconds. Cladding melting would begin after about 4.6 seconds.
The value at 15 seconds is used as the minimum time required for a local failure
to propagate to adjacent fuel modules. The response time of a thermocouple is
put at 1 to 2 seconds, with an unambiguous signal in 2 to 3 seconds. Estimates
put the system response time for detecting fission gasses due to cladding
failure at less than 5 seconds. This indicates that the detection of high
temperature and the shutdown of the fusion drawer before cladding melting will
be marginal. Sufficient time is available to prevent propagation of damage to
adjacent modules.

It is not clear if instrumentation would be required for each blanket
module. The latest LLL/GA design has monitors for each 12-module assembly.
The coolant manifolds then connect all portions of the blanket. Older designs
for the MHR called for segmenting the blanket into 16 isolated orange peel
segments, each with 45 modules. Although the 12 module assembly required
more instrumentation, it means that less modules have to be pulled and inspected
after the detection of activity or a flow disturbance. The use of the PCRV
eliminates the need to isolate the blanket into so many independent segments.

The operating characteristics of the mirror fusion driver will be similar
to the tokamak. However, the MHR will operate in a driven mode with the fusion
reaction maintained by beam injection. The fusion reaction can then be rapidly
quenched by stopping the beam injection. The relatively steady state operation
of the mirror driver will not result in thermal transients in the fission
bianket as in tokamak operation for qualification of fuel and cladding.

The potential for damage to the fission blanket in the event of a plasma
dump to the first wall is increased in the MHR since the fission modules them-
selves face the plasma. A burnthrough of the first wall would in fact consist
of holing the 2.0 mm pressure shell(s) of a module (or modules). The resulting

X-21
depressurization would quench the fusion reaction, but the resulting pressure
and thermal shocks should be investigated for producing fuel damage. Indica-
tions are that all structures can withstand the resulting helium pressure
transient.

If fuel damage should occur in the MHR, the potential for propagation
to nearby modules and the severity of damage will 1ikely depend on the
location of the initial failure. Fuel melting with failure of modules
in the upper portion of the blanket may result in widespread contamination in
the vacuum system with debris falling on modules located below. The prop-
agation of damage may then be far removed from the site of initial failure.
Locating the fuel modules directly facing the plasma rules out the use of
guard vessels to contain the spread of contamination.

b. Tritium Safety

The LLL/GA MHR design calls for the use of 1ithium deuteride (LiD)
pins clad in Lockalloy 43. This aluminum-beryllium alloy was chosen
especially for its low tritium permeability. The LiD also has a high deuterium
vapor pressure, making this a good choice of material for a batch processing
method of collecting bred tritium. No online tritium extraction process
would be required with this design, although a clean-up system would still be
needed. However, a subsequent economic analysis in the LLL/GA report states
that the costs and tritium availability associated with batch processing
were unacceptable. It was then suggested that a 1ithium compound which pro-
motes dehydriding be coupled with a relatively permeable cladding for online
extraction of tritium from the reactor coolant. The design considered in
this report was 1ithium oxide (Lizo) clad in stainless steel.

Going to an online method of tritium extraction will have a significant
impact on required safety cleanup systems. A large fraction of the tritium
inventory then becomes available for release to reactor containment in the
event of a depressurization accident, and the cleanup systems must be
designed accordingly.

Release to the environment was put at ~I0 Ci/day; due to losses into the
cooling systems of the main reactor, the neutral tritium beam injector and the

X-22
direct convertor. After extraction by cleanup systems, permeation of tritium
into the steam generators with subsequent loss to the environment resulted

in ~3 Ci/day from each of these sources. It is likely then that routine
tritium releases can be held to ~10 Ci/day for either online extraction or
batch processing of tritium.

c. Lithium Safety

No T1iquid 1ithium is used in the mirror hybrid design. Accident analysis
would then be simplified to examining the credibility of scenarios capable
of producing 1iquid 1ithium metal which could interact with the concrete of the
surrounding PCRV structure. It is likely that in postulated accidents this
energetic, damage by 1ithium reactions would be insignificant in comparison.

d. Magnet Safety

The implications on magnet safety and exposure to magnetic fields and
required research have been discussed for the tokamak. As already mentioned,
the introduction of the PCRV in the mirror hybrid is a major safety advantage
in that it virtually eliminates the potential for energetic magnet failure
leading to damage of the fission blanket structure.

For magnet repair, it appears the upper coil can be removed remotely.
Repair or rewinding operations must then be examined in light of activation
of the niobium, tin, and copper materials used. Removal of the bottom
ccil appears to be very difficult, and would be attempted only if the coil
required rewinding. Repair would otherwise take place in the end tank of
the direct convertor. This will require a safety evaluation of radiation
fields in this region, and likely require portable shielding for personnel.

X-23
E. THETA PINCH

1. Description of the Theta Pinch Hybrid Reactor Concept

The description given here is based on the Los Alamos Linear Theta

Pinch Hybrid design.(s)

This design consists of a cylindrical plasma chamber
20 cm in radius and 500 meters long. The actual reactor is divided up into
200 modules, each 2.5 meters in length. An insulator (graphite) lines

the plasma chamber and separates the shock implosion heating coil from the
return current generated in the gas when the coil is fired. A multiturn
adiabatic compression coil surrounds the implosion coil. The fission blanket
then consists of fuel assemblies oriented along the plasma axis in four radial
zones followed by a reflector, with the helium cooled Tithium region just
outside of the coils. No biological shields are placed around the reactor
itself. The entire device is placed within a steel-lined linear trench

which serves as the vacuum vessel and also provides containment in case of
accidental release of radionuclides. Concrete surrounding this vessel
provides structural support and acts as a biological shield. Penetrations
are provided for the helium delivery and return ducts every 2.5 meters, with
the main helium manifolds outside of the vacuum vessel.

The capacitor banks for the implosion heating coils are located just
outside of the concrete walls to the linear trench. Homopolar generators
are used for the compression coils, and are also located outside of the trench.

2. Safety and Licensing Issues for the Linear Theta Pinch Hybrid Reactor

a. Operating Characteristics

The LTPHR is based on another magnetic confinement fusion driver, but
where the tokamak and mirror examined previously operated in a quasi-steady
state mode, the theta pinch is a pulsed device. The burn time for the fusion
neutron source is 10 milliseconds with this particular design firing several
times per second (2.3 Hz). Depending on the exact fusion performance and
energy multiplication in the blanket, this device will then likely produce
on the order of ~10 megajoules of energy per pulse per meter of length.

With this mode of operation, the qualification of fuel materials and cladding
over the lifetime of the fuel cycle must be established for licensing.
Startup procedures for bringing the fusion driver up to power with a cold

X-24
blanket must also be established. It is unlikely that amy licensed perform-
ance envelope will allow initial full fusion driver output with high-energy
multiplication in the fission blanket.

b. LOCA/LOFA Analysis

The LTPHR is again a helium-cooled design. The geometry of the theta
pinch is such that a more conventional fuel lattice can be designed which
accepts fuel elements similar to those used in HTGRS. Analysis of local
flow blockage accidents may then closely parallel modern gas-cooled fission
reactor experience. However, the 500 m length of the reactor will likely
introduce special design requirements to guarantee the integrity of cooling
and vacuum systems under the influence of large external forces such as

earthquakes. A number of independent cooling systems along the length of
the device should be used.

In the event of fuel melting, the liner geometry of the theta pinch
again is ideal to virtually eliminate the chance of a critical reconfiguration.
The steel lined linear trench can be designed to contain any accidental
releases and should allow for relatively easy cleanup and decontamination.
The inlet parts to the vacuum system should be relocated away from the bottom
of the Tined trench and equipped with valves to eliminate the potential
for contamination of the vacuum pumping network in the event of a release of
volatile fission products.

The steel Tiner can also act as a primary tritium barrier by cooling the
surface or applying coatings to lower permeability. Thermal insulation
would, of course, surround the modules to reduce heat flow te the liner walls.

c. Coil Safety

The geometry of the theta pinch allows for a compact, easily accessable
fission blanket. However, it also places the fission assemblies in very
close proximity to the heating and compression coils compared to the
tokamak or mirror. In this design, the input energy to the shock coil is
0.37 MJ/m per pulse, with 33.4 MJ/m going to the compression coil. With the
coils placed between the plasma and blanket, any protective barriers designed
to restrain damage in the event of coil failure will impact the blanket

X-25
neutronics. A safety/performance analysis of coil failure mechanisms and
required engineered safeguards and diagnostics is then required. Failure
modes will 1ikely result from the cyclic nature of operation rather than
stresses beyond the design limit. This is because energy is delivered from a

fixed storage supply.

Accident scenarios involving loss of insulation at the first wall should
also be examined for potential damage to coils. The resulting electric

arcing could damage coils directly or lead to depressurization with mechanical
failure of components.

Associated with coil safety will be the safety of the pulsed power
supplies. There are advantages in locating the capacitor banks close to the
coils, which would place them just outside the concrete biological shield
of the trench containing the reactor. However, this will also be the likely
Tocation of the main helium delivery manifolds and cable trays for reactor
diagnostics. If the capacitors or homopolar generators are subject to
energetic failure, appropriate engineered safety barriers and damage control
systems will be required. The helium manifolds will also contain a circu-
lating fission gas inventory due to expected fuel failure, which will make
additional shielding necessary if maintenance will be required on electrical
systems located nearby.

X-26
F. LASER FUSION HYBRID

1. Description of the Laser Hybrid Reactor Concept

The reactor description given here is based on the Lawrence Livermore/

(7)

where laser irradiation of small targets containing deuterium and tritium

Bechtel design. This concept uses an inertial confinement fusion driver,
yield fusion neutrons. In this design, 1 MJ of laser energy on target

yields 100 MJ, with the fusion "microexplosions” confined to a cylindrical
vacuum vessel 10 meters in diameter and 16 meters high. The cylinder is
capped with upper and lower tritium breeding regions containing lithium

clad in 316 stainless steel with beryllium and graphite added. These regions,
along with the first wall, are cooled with liquid 1ithium.

The fission blanket is in eight sections placed around the cylinder in
three rows of hexagonal stainless steel fuel elements. A 19-rod cluster of
wire-wrapped stainless steel clad fuel pins is contained within each element.
The fission blanket is liquid sodium cooled with upper and lower 1iquid sodium
plenums capping off the blanket.

Finally, the cylindrical fissioh blanket is surrounded by a third lithium
zone, also liquid T1ithium cooled.

The reactor is designed for easy access and replacement of fuel assem-
blies or structural materials. The cylinder top can be removed, along with
the attached upper 1ithium blanket and first wall. Access to the bottom
lTithium blanket requires the removal of two segments of the radial fission
blanket. The top pleuum covers for the fission blanket are removable for
easy access to the fuel elements.

The fuel elements themselves resemble current fission fuel elements;
however, they are large by comparison. The Tength is put at 9.8 meters,
weighing approximately 1500 kg with 1200 kg of fuel.

As with coolant penetrations in LWRs, all coolant piping with the laser
hybrid enter and exit at one plane at the top of the cylindrical barrel,
The reactor is supported at the mid-plane.

X-27
A selenium laser system is used, present in the form of carbonyl selenide
(COSe). Electron beam excitation of xenon is used to disassociate the CQSe
molecule and excite the laser atom.

2. Safety and Licensing Issues for the Laser Hybrid Reactor

a. Operating Characteristics

The laser fusion concept also operates in a pulsed mode, with the pulse
rate varying from 8.5 to 5.5 cycles per second over the fuel exposure. A
major licensing effort will then be directed towards qualifying materials,
reactor systems and fuel assemblies in this nuclear environment. The radiation
damage problem from the 100 MJ microexplosions is expected to be severe,
with standoff considerations the reason for the large cavity diameter.
With this design, the first wall structure (1 cm thick graphite blocks brazed
onto a 1 mm molybdenum backing) is replaced every 1.5 full power reactor
years. The top blanket is also replaced at this time. After 3.0 full power
years it is estimated that all 8 segments of the reactor will need replacement
due to neutron damage, including top and bottom plenums. Materials perform-
ance then helps define waste production in addition to structural requirements
and operating procedures.

For accessibility and maintenance, the laser hybrid blanket incorporates
a number of positive design characteristics:

° unit fabrication and installation

o coolant piping entry and exit at a central plane

° removal capability of the total or part of the core without welding
or cutting

° easy access to fission fuel process tubes

In the event of an accident or malfunction, the fusion driver can be cut
off simply by shutting off the laser.

b. LOFA/LOCA Analysis

The laser hybrid concept has a significant advantage in accident analysis
over the other fusion drivers in that the high energy systems used to initiate
the fusion reaction are removed from the vicinity of the fission blanket.

X-28
The laser facility itself is located in another building along with its
power supplies, with the beams transported through underground tunnels. To
insure the integrity of the reactor containment, a series of fast acting
valves can seal off the tunnels in accident situations.

The considerations for initiating events for loss of flow or loss of
coolant type accidents then becomes essentially those for an LMFBR facility.
The design further includes features to mitigate .the consequences of potential
LOFA/LOCA scenarios:

o The fission blanket is subcritical in all configurations.

° The fuel elements in the reactor are furnished with a diaphragm that
serves as a secondary containment to prevent loss of coolant. The
problem can also be isolated due to the independent modular blanket
design.

A11 equipment and piping containing lithium or sodium is housed in
steel lined vaults containing an inert gas.

Again, the modular hybrid fission blanket makes it possible to isolate
coolant disturbances; however, the geometry and plumbing are more complex
than an LMFBR around the vessel. A loss of coolant accident due to pipe break
or leak in the vessel would probably cause some fuel melting and slumping.
The LLL/Bechtel report indicates that this type of accident may be difficult
to cope with due to the large size of the blanket structures surrounding the
vacuum vessel and the problem of surrounding all of the primary system com-
ponents (blanket plenums and process tubes) with guard vessels. No detailed
analysis has been performed. \

The major differences in the safety analysis will be due to the liquid
lithium inventory in this design, and the presence of the large vacuum
vessel. The safety research required for the liquid 1ithium was outlined in
the discussion of the tokamak hybrid. Activated corrosion product transport
in the liquid metal systems will require investigation.

c. Tritium

For tritium the goal is to limit release rates to 0.0021 grams (20 Ci)
per day, similar to PWRs. A full, in-depth analysis of tritium leakage rates
from fusion equipment and recovery systems has not been made, but design
estimates have been made. The primary coolant loop is designed to hold the

X-29
tritium vapor pressure at 10'8 torr, with diffusion into the secondary loop
put at 0.1 gram per hour before partial recovery. This should 1imit releases
to the steam system to 1 to 2 Ci per day. For accidental releases of tritium
in containment, a system capable of recovering a loss of 50% of the total
inventory (10 kg) is provided. The system operates at 150,000 cfm and is
designed to reduce airborne tritium levels to 5 uCi/m3 in less than three days.

d. Laser Safety

The Taser system itself introduces a number of safety issues. These include:
o laser beams

° laser power generation
° chemical processes

In this design, a selenium laser with electron beam excitation is used,
relying on capacitor banks for energy storage. The capacitor system will
require standard safety procedures for maintenance and to contain damage in
the event of failure. The e-beam source will require shielding or non-access
during operation to prevent exposure to X-rays. Current procedures for firing
e-beam excited CO2 lasers are to clear the laser hall, with personnel restricted
to the control room.

The active laser gas in this design, carbonyl selenide, is toxic.
Therefore, the laser system is leak tight, along with the laser building, which
operates at a pressure less than one atmosphere.

Provisions are also made to pump down and condense the carbonyl selenide
in the event of an accidental release to the building. A release of the entire
selenium inventory is estimated to bring concentrations to~200 times the
allowable limits. Methods of detecting leaks and monitoring airborne concen-
trations in the laser hall are then required. |

e. Fuel Handling

The geometry of the laser hybrid blanket and fuel assemblies most
resembles concepts used in pure fission reactors compared to the complex
geometry of fuel modules used in magnetic fusion driven hybrids. As such,

X-30
the safety analysis for access and fuel handTing will more closely resemble
procedures used and licensed in the fission industry. The only major differ-
ence will be in the size of the fuel assemblies used. Fuel handling machines
will have to be scaled up for the 9.8 meter length and 1500 kg mass.

X-31
G.

SECTION X REFERENCES

Tokamak Engineering Test Reactor, UWFDM-191, University of Wisconsin,
Department of Nuclear Engineering, Madison, WI 53706, June 1977.

D. J. Bender, et al., Reference Design for the Standard Mirror Hybrid
Reactor. UCRL-52478, General Atomics and Lawrence Livermore, Livermore,
CA, May 1978.

Proposed ANS Standard Decay Energy Release Rates Following Shutdown of
Uranium-Fueled Thermal Reactors, ANS-5-1, American Nuclear Society,
Chicago, IL, October 1971.

S. F. Su, G, L. Woodruff and H. J. McCormick, "A High-Gain Fusion-Fission
Reactor for Producing 233U," Nucl. Tech. 24:392, June 1976.

Environmental Development Plan (EDP), Magnetic Fusion Program.
DOE/EDP-0008, U.S. Department of Energy, Washington, D.C., March 1978.

R. A. Krakowski, et al., "An Engineering Design of a Linear Theta-Pinch
Hybrid Reactor (LTPHR)," TANSA 21:61 (1975).

Laser Fusion-Fission Reactor Systems Study. P0-5040603, Bechtel
Corporation, July 1977.

X-32
XI. ENVIRONMENTAL CONSIDERATIONS

A fusion-fission hybrid reactor is expected to be a large-scale thermal
energy facility. As such, many of the environmental impacts associated with
the concept will be similar to those of modern fission reactors. This includes
site selection and many of the impacts of construction (site work, materials
requirements, influx in population, etc.). In this section a generic discus-
sion of the hybrid concept will be oriented towards identifying any unique
environmental impacts.

A. FUSION FUEL CYCLE

1. Deuterium and Lithium

A11 fusion drivers in this report are based on the deuterium-tritium fusion
reaction. The fusion fuel cycle then requires that the basic materials of
deuterium and 1ithium be delivered to the nlant. A large lithium inventory
(possibly in 1iquid metal form) surrounding the fusion reaction is then the
target material for creating tritium by neutron capture.

The procurement of deuterium is expected to be routine. Deuterium occurs
in all natural waters at a concentration of about 150 opm, and the world

13 metric tons. It currently is readily

inventory is estimated to be about 10
extracted and is available commercially at a relatively low cost ($600/kg).
Since at least a quarter of the current resources can be extracted without a
significant increase in cost, an essentially unlimited supply is available at

current costs.

The deuterium is obtained from water by use of a hydrogen sulfide extraction
process (the Guerdler-Sulfide or G-S process) to obtain heavy water (DZO) and
then electrolytic decomposition of the heavy water to obtain the deuterium.

This process has been used commercially on a large scale for over 20 years and
has an insignificant environmental impact consisting primarily of processing
small quantities of water and releases of very small amounts of H,S and SO,.

Several hundred tons of T1ithium are typically used in the blanket of fusion
reactors, and hybrids will require similar amounts. Only about 1% of the inven-
tory will be consumed during a 30-year lifetime of the power plant in breeding

XI-1
tritium. Since this small consumption is less than the extra amount of 1ithium
that would be kept in storage as an emergency sunply, it is probable that no
additional shipments of 1ithium would be required beyond the initial startup.

The original procurement of Tithium will require mining, milling and
processing operations. This will impact land and water use, and produce waste
piles, waste ponds and chemical releases to the water and air. It is expected
that modern waste control technology can prevent any serious adverse impacts.

2. Tritium

As with safety and licensing, the primary environmental concerns will
center on the potential for routine and accidental release of radioactive
materials. Tritium will 1ikely be the dominant radionuclide released,
present in solid, liquid and gaseous effluents. A preliminary evaluation of
the performance of radwaste systems indicates that even though tritium will
be present in kilogram quantities in reactor systems, the routine release to the
environment can be kept to levels found in light water fission reactors (<20 Ci/
day).

The consequences of an accidental release of a large tritium inventory
must also be addressed. The worst credible accident in this regard is con-
sidered to be the failure of a liquid Tlithium blanket with a successive failure
of the fire suppressant device. An analysis of a laser fusion reactor accident(])
led to a maximum dose at the side boundary (100 m) of 0.7 rem for a cool
ground level release, and 7 x 10'4
would reduce the latter by a factor of 100. The tritium inventory in a hybrid

rem for a hot fire release. A 100 m stack

reactor would be less, resulting in a smaller release. To put the biological
hazard of tritium in perspective in this worst possible accident, it was noted
that someone at the site boundary would receive a fatal chemical exposure to
the Tithium smoke long before one could receive a lethal tritium dose.

Tritium will also likely contaminate solid structures removed from the
reactor for maintenance. Any solid waste disposal must then be examined for
gradual tritium leakage to the environment.

Since tritium is expected to be the primary cause of radiation doses to
the general public as a result of radioisotope releases, ample technology must
be available for estimating the release rates, radiation doses and biological
effects. The following information is needed to assure adequate ability to
write environmental statements:(z)

e Tritium permeation rates through fusion reactor structural materials.

e Tritium separations chemistry, including chemical and physical
equilibrium relationships.

e Tritium separation processes for air and water streams.

e Optimum tritium storage methods.

o Tritium barrier technology.

e Application of current tritium control technology.

e Detailed designs for power plant subsystems containing tritium.

e Tritium concentrations and doses at Tong distances from release points.
e Tritium transport through.the biosphere.

e Additional information on the relationship between dose and somatic
and genetic effects, especially in relation to long-term exposure to
tritium at very low concentrations.

3. Activation Products

The D-T fusion fuel cycle will produce an intense neutron source, leading
to activation of structural materials and coolant impurities. Neutron streaming
from ducts and beam ports can also lead to substantial generation of activa-
tion products. All of these must then be examined for source terms into the
environment.

The replacement of structural materials due to radiation damage is expected
tc generate the bulk of the solid radwastes. Corrosion and erosion in the
coolant system and vecuum system may also lead to waste streams which must be
packaged and disposed of. The activated structural materials removed from the
reactor are not thought to present any hazard in terms of an accidental dis-
persion into the environment, although they will require shielding. The
materials will be removed in such quantities that they will represent valuable
resources, and it is likely that they will be stored for recycling after a
period of radioactive decay.

XI-3
While still in the reactor, only the most energetic accidents postulated
would be capable of releasing radioruclides to reactor containment and then
possibly to the environment. Liquid metal fires or melting of fuel assemblies
in the fission blanket are considered to be the only plausible methods of
releasing the activation products in significant quantities; however, these
accident scenarios would likely have more serious consequences than those associ-
ated with the release of activation products.

B. FISSION FUEL CYCLE

The four hybrid fuel cycles investigated in this report are as follows:

1. Once-Throuah - natural uranium fueled hybrid in throwaway mode (power
production only).

2. Pu-Recycle to Thermal Reactors - hybrids with dual role of fissile
fuel production and power production.

3. Refresh Fuel Cycle - hybrid reactor re-enriching spent PWR fuel for
return to PWR.

4. Pu-Th (Pu Catalyst) Fuel Cycle - hybrid reactor breeds 233U in

plutonium-thorium target; 233U sold while the plutonium is recycled.

The environmental impacts associated with these fuel cycles will then
come from acquisition of materials, initial fabrication, transportation, operation
in the hybrid, transportation of spent fuel, reprocessina and waste storage.
Note that none of the fuel cycles being considered require enrichment of the
original uranium feedstock, thereby eliminating impacts associated with gaseous
or centrifuge enrichment plants.

The environmental impacts associated with acquisition of uranium and
thorium and initial fabrication are identical to those now experienced with
1ight water reactors. Hoviever, the resource utilization, or power generated
per metric ton of ore mined varies widely with the fuel cycles considered.
The initial material requirements and mass flow diagrams in Chapter 6.B.
indicate that the fuel breeders are capable of supporting several fission
reactors with their fissile fuel production; however, a once-through hybrid
is a very inefficient use of natural uranium. Due to the relatively low
thermal power densities in the hybrid blanket, this throwaway fuel cycle requires

XI-4
a much higher uranium supply per kWe generated as compared to light water
reactors. In order of most efficient use of natural resources, the fuel cycles
are then Pu catalyst, Pu recycle, refresh, and finally, the once-through,

The fission products and actinides in the fission blanket will 1ikely be
relatively benign in routine operation of the hybrid reactor, as is the case
in pure fission reactors. However, they have the potential for causing the
most serious environmental damage if accidentally released at some point in
the fuel cycle. Because of this, the hybrid reactor will require the same
safeguards in reactor cooling, containment and aerosol blowdown systems.

Of interest here is the consideration that the fusion driver may produce
a unique fission product inventory in the hybrid blanket. The neutron spectrum
in a hybrid reactor is very different from that in a thermal or "fast" fission
reactor due to the 14 MeV fusion neutron source and the subcritical nature of
the blanket. Very fast fission reactions will result in a different fission
yield than normally experienced, with the probability of symmetric fission
increasing by two orders of magnitude. The abundance of fission products with
atomic mass number between 105 and 130 in the hybrid will reflect this fact.
The actual distribution of fission products and actinide will further depend
on the geometry of the blanket and the particular fuel cycle used. The research
required for the identification of specific radionuclide inventories and the
perforrance of fuel systems in the hybrid were discussed previously in the
chapter on safety and licensing.

As with safety analysis, the verification of the fission fuel cladding
in the hybrid nuclear environment will be one of the more important require-
ments for licensing from an environmental standpoint. Although the public
tends to focus on the potential for large accidental releases from a nuclear
facility (which certainly must be evaluated), in actual practice the routine
release of small amounts of fission gases will determine the actual environmental

impact. It must then be established that the hybrid fission blanket and associ-
ated cleanup systems can routinely perform up to the standards set for the
fission industry in the nuclear environment of the fusion driver.

XI-5
However, it is the accumulation of the actinides, including plutonium,
that will have the greatest impact on the potential environmental hazard
presented by the hybrid fuel. The isotopes of americium, Am-241 and Am-243,
are of particular biological concern. The once-through, Pu recycle and
refresh fuel cycles have essentially no fissile fuel loading initially, but
as seen in the performance tables in Chapter VII.B, these fuel cycles all
produce plutonium in metric ton quantities per full power reactor year of
exposure. The build-up of other actinides such as americium is then dependent
on the neutron flux spectrum in each specific blanket design. The hybrid
has the potential for being an actinide burner, however EPRI studies(3)
indicate that the inventory of actinides increases significantly before con-
sumption by fission is effective.

The remaining fuel cycle, the Pu catalyst, starts with an initial
inventory of several metric tons of plutonium for all hybrids considered in
this report. This Pu inventory then drives the thorium-uranium scheme:

>

(4) suggested that higher actinides
for the thorium-uranium fuel cycle would only appear if the U-233 were left
for very long times in the blanket, thereby reducing the hazard potential
for the fuel cycle. However, the use of the plutonium catalyst puts this in
doubt. The buildup of U-232 which has a long decay chain of alpha emitters
may also present a problem. Again, the proper neutronics evaluation of
actinide buildup for the four fuel cycles is not available.

A previous examination of Th-U cycle

The concentration of plutonium in the hybrid fuel and its isotopic
composition must also be addressed. With the tokamak production rates in
Section VII, Pu equilibrium concentrations in the discharged fuel range from
.025 to .07 MT Pu/MT for the four fuel cycles. This compares to typical Pu
discharge concentrations in PWRs of ~0.01 MT Pu/MT (250 kg Pu in 1/4 core
discharge, 33,000 MWd/MT burnup). The isotopic composition of plutonium
in discharged PWR fuel is also typically spread over several isotopes (1.7%
Pu-~238, 55.8% Pu-239, 24.5% Pu-240, 13.1% Pu-241, 4.9% Pu-242) where the
plutonium in the hybrid fuel is expected to be >90% Pu-239.

XI-6
The presence of fission products in the hybrid blanket after exposure
has been addressed, however no specific radionuclide inventories are
available for inclusion in this report. A previous examination of this
prob]em(3) indicates that the change in the fission product yield curve
for 14 MeV fusion neutrons will result in a higher production rate (compared
to LWRs) of hazardous radionuclides such as ruthenium-106. However, burn-
up reactions such as (n,2n) may possibly reduce the difference in fission
product inventories to insignificant levels. Fission product inventories
in specific blanket volumes must, of course, be scaled to blanket power
densities (250-500 watts/cm3 for LWRs) and exposure. The proper neutronics
must be developed for the fusion neutron spectrum if these fuel cycles are
to be investigated properly for licensing.

It is then likely that per unit of power produced, the environmental
impact of the fission products in the hybrid fuel cycles will be similar
to those now experienced in light water reactors.

The environmental impact of the fission fuel cycle must also address
the potential for release of these materials during transportation and
reprocessing, if any. The technology employed is expected to be identical
to that now developed for the fission industry. The analysis of transporta-
tion accidents must then address the higher concentrations of plutonium and
other actinides in the spent hybrid fuel. The effluents released during
reprocessing for all fuel cycles except the once-through are assumed to be
those of their pure fission counterparts. Again, the neutronics are not
available to estimate the release of the radioactive noble gases, krypton and
Xenon.

For high level waste storage, the once-through hybrid fuel cycle will
require the disposal of metric ton quantities of plutonium each year. This
is in addition to actinides and fission products. This will 1ikely be unac-
ceptable from a resource utilization and waste management point of view. The
environmental impact analysis must address the relative hazards associated with
long term storage of spent fuel with these high fissile material concentrations
as opposed to reprocessing.

XI-7
C. MAGNETIC FIELDS

Three of the fusion drivers in this report are based on the magnetic
confinement of the fusion reaction. The safety aspects associated with
occupational exposure and the research required were outlined in the chapter
on safety and licensing.

Where the safety aspects of magnetic fields were concerned with
occupational exposure to field strengths as high as several hundred gauss
for short periods and several tens of gauss for long periods, the environ-
mental impact will be determined by exposure to field strengths comparable
to the earth's (~0.5 gauss). For example, with the tokamak driver in this
report, the toroidal field strength is over 60 kilogauss in the plane of
the torus, but it drops off rapidly and will be far below 0.5 gauss at the
site exclusion boundary (800 m). However, the poloidal field radiates both
horizontally and vertically from the torus, and is expected to present a

public exposure similar to that from the earth's natural fie]d.(G)

The ability to demonstrate conclusively any biological effects associated
with exposure to high magnetic field strengths is proving to be a difficult
enough research task. Because of this, it is unlikely that any direct
demonstration of effects from exposure to field strengths on the order of
1 gauss will be possible. In all probability, the environmental impact
assumed for licensing purposes will be based on some extrapolation of effects
observed (if any) at high exposure levels. This requires that a non-
threshold assumption be made similar to that used with low level ionizing
radiation exposure.

D. TOXIC LASER GASES

In the laser fusion driver used in this study, the active laser gas,
carbonyl selenide, is toxic. Accordingly, the laser building operates at
a pressure of less than one atmosphere to contain routine releases.
Provisions are also made to pump down and condense the carbonyl selenide in
the event of an accidental release to the building. A release of the entire
selenium inventory is estimated to bring concentrations to ~200 times the
allowable T1imits. Because of the above precautions the operation of the
laser driver is not expected to have any measurable impact on the outside

X1-8
environment even under the worst accident scenarios. However, the shipment
of carbonyl selenide to the reactor site will have environmental impacts
similar to the shipment of other toxic gases.

E. UNIQUE RESOURCE REQUIREMENT

A Tlarge hybrid reactor economy can possibly result in a significant
increase in demand for materials associated with the fusion driver, including
the Tithium discussed earlier. These include the possible use of beryllium
as a neutron multiplier, or materials chosen for their resistance to radiation
damage and reduced neutron activation in the intense fusion nuclear environment.
The particular designs in this report use stainless steel as a structural
and fuel cladding material.

The impact on domestic demand and materials supply are shown in Tables
XI-E-1 and -2 assuming a ccmmercial fusion economy with 2810 GWe insta]led.(z)
The materials consumption in the hybrid should be lower per GWe installed
due to the significant energy production in the fission blanket.
TABLE XI-E-1. Total Domestic Demand for Important Fusion Materia]s(z)
(1975 to 2040)

Without With
Units Fusion Reactors Fusion Reactors
Beryllium 10° metric tons 300 2,440
Chromium 108 metric tons 140 180
Copper 106 metric tons 410 440
Iron Ore 109 metric tons 25 25
Helium 109 cu meters 4 7
Mercury 106 35-kg flasks 16 16
Lithium 102 metric tons 810 5,960
Molybdenum 106 metric tons 6 8
Nickel 106 metric tons 50 80
Lead 106 metric tons 690 760

TABLE XI-E-2. Depletion of 1974 U. S. Reserves of Important
Fusion Power Plant Matgria]s, Assuming No
Additions to Reserves?

U.5. Reserve Depletion Date

Without With

Material Fusion Reactors Fusion Reactors
Beryllium Before 2010 Before 2010
Chromium --Al1 chromium currently imported--
Copper Before 2010 Before 2010
[ron Before 2010 Before 2010
Helium After 2040 2030
Mercury Before 2010 Before 2010
Lithium After 2040 2020
Molybdenum 2030 ‘ 2,2020
Nickel Before 2010 Before 2010
Lead Before 2010 Before 2010

XI-10
SECTION XI REFERENCES

J. J. Devaney, L. A. Booth, J. H. Pendergrass and T. G. Frank,
Environmental Effects and Potential Hazards of Laser Fusion Electric
Generating Stations. LA-UR-77-2275, presented at ICF Conference,
February 7-9, 1978, San Diego, CA.

J. R. Young, An Environmental Analysis of Fusion Power To Determine
Related R&D Needs. BNWL-2010, Pacific Northwest Laboratory, Richland,
WA, November 1976.

Fusion-Driven Actinide Burner Design Study, EPRI-ER-451, Electric Power
Research Institute, San Diego, CA, May 1977.

D. J. Bender, et al., Reference Design for the Standard Mirror Hybrid
Reactor, UCRL-52478, GA-A14796, General Atomic Company and Lawrence
Livermore Laboratory, Livermore, CA, May 22, 1978.

W. E. Kastenberg, et al., Some Safety Studies for Conceptual
Fusion-Fission Hybrid Reactors. University of California, Los Angeles,
CA 90024.

C. E. Easterly, K. E. Shank, and R. L. Shoup, "Radiological and
Environmental Aspects of Fusion Power," Nuclear Safety, 18:56,
March-April 1977.

XI-1
XIT. UTILITY AND INDUSTRIAL PERSPECTIVES - COMMERCIALIZING HYBRID REACTORS

Technological change occurs when it becomes possible to employ a new
technique, such as fusion-fussion (hybrid) reactors in the production of goods
and services. The extent to which a new technique is adopted is generally
dependent upon three economic considerations: direct costs to the users,
extra market costs to users and non-users alike, and the efficiency of the
market(s) for the new technique. The direct costs of hybrid reactors which
are of concern at this point in time are the maximum allowable capitalized
costs permitting the technology to penetrate electric generation markets.
These have been considered in Section IX. Even if the technology is poten-
tially cost effective, it still may not become a commercial success for a
variety of reasons. Further, in establishing public policy it is necessary
to consider all costs of employing a production technique. In this section we
address the economic issues related to the extra-market costs and market devel-
opment and efficiency for an emerging technology such as fusion-fission.

In the following paragraphs we shall address the utility and industrial
perspectives on hybrid reactors within the context of the commercialization
process. This will include a statement of the scope and general theory of
commercialization. From that foundation, specific issues in the process can
be identified and reviewed for the case of hybrid reactor concepts. The objec-
tive is to illuminate the key factors which will influence private sector's
decisions to invest in fusion-fission reactors. In turn, some of the public
decision making problems will be highlighted.

A. SIGNIFICANCE OF COMMERCIALIZATION ISSUES

In recent years the Federal Government has allocated a substantial
proportion of its resources to civilian research and development activi-
ties. The Department of Energy and a predecessor organization, the Energy
Research and Development Admininstration, exemplify this trend. The objec-
tive of these activities has been to hasten the development of new tech-
nologies and smooth the transition from one technology to another. For these
research activities to be effective, viable technologies must be integrated

XI1-1
into the economic system. That process has become known as "commercializa-
tion". The proper management of research and development will recognize
and take into account this process. Failure to do so may result in limited
utilization of federal civilian R&D output. This can constitute a waste

of physical and intellectual resources if viable technologies are passed
over or inferior ones are forced on the market.

The process of commercialization has not been extensively researched
or even consistently defined.(s’a) Research to date indicates that commer-
cialization should be viewed as a process that begins early in the R&D pro-
cess. The specific timing and degree of involvement of commercialization
in R&D activities is not currently known; however, it is evident that
there is no prescribtion for commercializing a new technology.

The process of commercialization is essentially a matter of "market
development" and can, using familiar terms, be either related to "demand
pull" or "technology push". This "market development" process parallels the
technological development process in that both of these processes are striv-
ing to reduce uncertainty through the generation of improved information. As
shown in Figure XII-A.1, the process of commercialization has two distinct
phases, depending on the existence of a functioning market. 1In the early
stage, labeled market identification, the technology is not fully developed
and thus market transactions are not occurring. However, even at this time
a "psuedo market" exists where information concerning economic and technical
feasiblity is exchanged. The second stage is initiated by the introduction
of the new technology into the marketplace. Now actual market exchanges
involving the technology can occur, with the information flow continuing as
inferior products are "weeded out" and surviving products are continually
refined. As the process evolves, the level of information is increased with
a corresponding reduction in uncertainty.

XII-2
- Market —— Market ey

Technological Identification Penetration
Milestones
A A A A
Innovation Pilot Demonstration Commercial Scale
Plant

FIGURE XII-A.1. Scope of Commercialization

A similar argument has been advanced concerning industrial innovation:

. there are two sources of ambiguity about the relevance
of any particular program of research and development--target
uncertainty and technical uncertainty."(1)

The traditional view is that the reduction of "technical uncertainties" is
the principal objective of the R&D program. Today's concern for the rapid
development of new technologies has accentuated the notion of target uncer-
tainty. In the earliest stages of development, target uncertainty may
include both the vendors and consumers of a technology. The process of
commercialization can thus be viewed as the reduction of target uncertain-
ties. One can, of course, employ a broad definition of the target concept
for it includes many individual attributes of the ultimate market. Let us
then proceed with a more detailed analysis of the commercialization concept.

B. CONCEPTUAL MODEL OF THE COMMERCIALIZATION PROCESS

Commercialization can be viewed as a multi-dimensional process, com-
posed of four basic elements: market demand, property rights, capacity to
produce, and technological attributes. A full understanding of commerciali-
zation requires an integration of all of these. The primary interrelation-
ships among the elements are shown below in Figure XII-B.1. The elements
of market demand and capacity to produce constitute the primary components
of a market--a demand sector and a supply sector. The elements of property
rights and technological attributes represent the institutional and struc-
tural factors which determine the efficacy and efficiency of the market.

The logical starting place in this conceptual model of commerciali-
zation is with the effective market demand. Market demand is derived from
the wants and incomes of potential buyers. In thecase of hybrid reactors,

XII-3
Market
Demand

Property
Rights

Capacity to
Produce

FIGURE XII-B.1. Conceptual Model of Commercialization

demand is derived from the demand for goods which require electricity as

an input. Commercial acceptance inevitably depends upon the ultimate user's
willingness to pay for an R&D product, either directly or indirectly through
the purchase of some other good. The element of market demand is a primary
component of the theoretical literature on the determinants of technologi-

(2)

cal change.

For example, the theory of induced technological change is fundamen-
tally demand-driven.'3) Additionally, a body of applied 1iterature con-
cerning research utilization, new product development, and market research
typically involves identifying and satisfying needs as expressed in the

marketplace.

Market demand works through the institutional arrangement of property
rights to create incentives to produce. For a firm to undertake an invest-
ment to develop and produce a new product, the firm must be reasonably
assured of recovering its investment. If the firm cannot establish and
enforce rights to its product, that investment may not be forthcoming. At
this point, this is the principal argument for the establishment of patent
rights.

XII-4
The property rights of consumers also influence market demand. Some
types of goods, typically known as "public goods", are not subject to the
principle of excludability in use. Because individuals cannot be prevented
from consuming public goods once they have been produced, consumers will be
unwilling to pay private producers of such goods. Since private producers
cannot obtain complete compensation for their production, private markets
will fail to fully reflect the demand for public goods. Thus, R&D products
having attributes characteristics of public goods may have limited commer-
cial potential unless corrective action is taken by public agencies.

The establishment of incentives to produce through the institution of
property rights leads to the third element of the process, the capacity to
produce. Two issues are associated with this element. The first is the
problem of technological transfer which has arisen in the context of devel-
oped and developing countries, but is also relevant to the flow of infor-
mation between the research and production segments of an economy. Technol-

(4)

ogy transfer is vital to the commercialization process. The second issue

is related to the behavior of the producing sector and has been addressed

(3)

characteristics can influence a firm's decision to enter a new market or

in the literature on market structure and innovation. Market structure
adopt a new production technology. Thus, structural characteristics must
be considered in formulating commercialization policies.

The final element of the process is the technology's characteristics
as they relate both to market demands and to the producing sector. The
attributes of a technology must meet the basic needs of the consuming sec-
tor to be commercially viable. In managing R&D activities, information on
the market demands should guide the development of a new product's charac-
teristics. In addition, attributes of the technology influence the struc-
ture of the producing sector, and thus its conduct and performance. Such
features as product differentiation, economies of scale and economies of

(5)

scope are important in determining market structure.

In the following sections each of the four elements will be examined
in greater depth and implications for hybrid reactors will be considered.

XII-5
Through this framework the salient utility and industrial issues can be
identified. However, the scope of the current study is such that this
will provide only a first glance at the commercialization issues.
Several will merit additional in-depth analysis.

C. CHARACTERISTICS OF DEMAND

The primary function of any economic system is the satisfaction of
society's wants with the least cost combination of resources. Thus, the
economic concept of demand is determined by the relative prices of commod-
ities, the decision maker's budget and their preferences. The satisfaction
of these demands necessitates a set of technologies to produce goods and
services with the desired attributes. This establishes two principal ob-
jectives of new technologies. First, that they offer a more preferred
set of attributes for their cost. Second, the same attribute can be achieved
with fewer resources. In more familiar terms the first are product improve-
ment and the second are cost reduction innovations.

Recent research on the process of innovation has identified a pattern
related to the two principal motives for technical change with interesting
implications for the general process of commercialization. Abernathy and
Utterback(1) observed a generic pattern innovation beainning with new pro-
duct development as the technology matures the thrust of technological
changes are toward process innovations in order to produce the product more
efficiently. They also found that new products are often the results of
efforts by small technology-based companies while process innovations are
often made by the Targe manufacturing firms. First, the cost of change is
an increasing function of the size and integration of the organization.
Second, the user's input into the research process is vital to the success
of the innovation. Thus, the small organization can be more responsive to
the needs of the users and it is less costly for them to make radical changes
in product characteristics. However, as a technology matures there tends
to be a shift in the nature of competition from product characteristics to
price and cost sharing innovations are necessary for the organization's sur-
vival. The manufacturing firm is itself the user of the new technology.

XI1I-6
This has potential implications for the course of commercialization
for technologies developed by the Federal Government such as hybrid reactors.
Firms with substantial interest in producing the prevailing products would
at first g1an¢e appear to be logical producers of the next generation. How-
ever, this may not be the case; the cost of change to those organizations
may vary a great deal due to their specialized knowledge and operating sys-
tem. Therefore, entry with a new technology may likely come from smaller
firms not presently engaged in the market for the prevailing product. By
way of an example, the development and introduction of high temperature
gas-cooled reactor (HTGR) would appear to fit this pattern. Although Gen-
eral Atomic was unsuccessful in its first attempt to capture a share of the
electric generating reactor market, its attempt is noteworthy for other
advanced reactor concepts and should be considered in more depth.

The derived demand for hybrid reactors is primarily for their capacity
to breed fissile fuel. Thus, their demand will be influenced greatly by
the price of uranium. The future of the natural uranium market is at the
moment subject to considerable uncertainty and specu1ation.(6)' The uncer-
tainty itself suggests a need for the development of a 'breeder technology'.
Some industry representatives have implied that utilities will be willing
to build conventional nuclear reactors without a thirty year supply of
uranium provided fusion scientific feasibility has been demonstrated and
the Federal Government is actively supporting a hybrid reactor research
program.(7) The uncertainty in the market is seen in wide ranges of
price forecasts reported within the 1iterature.(7) Some suggest the

price will remain about $40 per pound (in constant dollars) through

the year 2000 and others see prices rising to more than $74 by then. The
recently published draft report on uranium from the Committee on Nuclear

and Alternative Energy Sources (CONAES) expressed a very pessimistic view

of the uranium reserves available given a price range of $40 to $60.(8) iIf a
high cost U308 scenario evolves there will be a demand for hybrids as fissile
fuel breeders.

A second component of the demand for hybrids will be their contribu-
tion to the development of pure fusion reactors. The timeframe for

XII-7
developing a hybrid reactor is perceived to be shorter than for a pure
fusion device due to the Tower requirements on the fusion component of the
reactor. There could be significant benefits derived from the learning
experience of operating a fusion-fission reactor which would carry over to
pure fusion designs. This would apply to the design and construction of
confinement systems and the fusion fuel cycle.

The development of a new technology is an investment process and,
therefore, dependent upon the function of the capital market. It is often
argued that new energy technologies fail to be developed because they can-
not attract sufficient capital and thus require government intervention.
This could either be a problem of insufficient return on the R&D invest-
ment or the inability to spread the risks. One could also find a problem
of under-investment in new energy technologies if the social discount rate
is less than the private sector's opportunity cost of capital. The private
sector decision makers could discount further benefits more than is socially
optimal. However, this would not be unique to the investment in energy
related technologies. There could be a differential impact if energy tech-
nologies are significantly more capital intensive or require a longer ges-
tation period for development than the typical product. Recent analysis
suggests that there is little reason to believe that the capital market cre-
ates a significant problem for commercializing new energy techno]ogies.(9

D. PROBLEMS OF PROPERTY RIGHTS

The role of property rights is central to the performance of a market
economy. The institutional arrangements of property rights serves as the
foundation for exchange relationships between those who demand and supply
various goods and services.éf An entrepreneur realizing the existence of
a demand will undertake an investment (in R&D and/or capital equipment)
necessary to supply the product provided he can earn a "normal” return on
his investment. If the entrepreneur's investment generates benefits ex-
ternal to his operation for which he is unable to obtain compensation,
then he will tend to under-invest in that activity. Because the output of

a/ For a complete discussion of the economics of property rights the
reader should see References 10 and 11.

X1I-8
R&D investments is information. for which it is very difficult to establish
and enforce property rights, there is a tendency to under-invest in research.

As was mentioned above the patent system was adopted to protect the returns
of the innovators. Sometimes patents are unenforcible or cannot be applied

to the given technology and thus offers another argument for direct public
support of innovative activities. The Department of Energy (DOE) generally
grants patent rights to contractors performing research for DOE, however,

they normally retain exclusive title. In this regard, there appears to be a
shift in DOE policy on patent rights which could become important for hybrid
development. In the area of coal technology and "synfuels", DOE has recently
granted foreign patent rights to a cost-sharing contractor for a coal lique-
faction process.(]z) With respect to fusion-fission reactors many feel that
the areas of blanket design and fuel cycles are ripe for patentable inventions.
An improved patent right policy could be important in generating significant

private sector involvement in the technology.

It should be pointed out that in other cases legal protection is
unnecessary. What have been termed "first-mover" effects create market
protection for the originator of the product or service. This can result
from two basic situations. One occurs when brand-name and identification
are important in the consumer's purchasing habits. The other occurs when
the lead times are very long for another firm to copy, produce, and market
a competing product. Due to the complexity of hybrid reactor concepts the
second factor could be important. There is already active interest in
hybrid concepts by several large private firms (i.e., Westinghouse, General
Atomic and Exxon through their research laboratory and their subsidiary
Exxon Nuclear). Involvement of such firms in technologies such as hybrids
is very important. However, if their interest is primarily research for
profit as opposed to becoming a vendor of the technology, then the signifi-
cance of their activities for commercialization is greatly reduced. It
would appear that a detailed investigation of role property rights in com-
mercializing fusion-fission reactors would be valuable in sorting out the

XII-9
potential for patentable inventions as well as the motives the first pri-
vate firms involved in the technology.

In addition to patent rights there is another institution poten-
tially useful in developing hybrid concepts, which can deal with the
ownership problems associated with research output. In most sectors, indus-
try wide research associations would be difficult if not impossible to es-
tablish and administer due to anti-trust considerations. However, this is
not a problem in the electric generating industry and so collective research
associations can aid in the development and commercialization processes.

The Electric Power Research Institute (EPRI) and a more specialized case,
the Gas Cooled Reactor Associates (GCRA), are two examples of these. Such
groups channel funds into research generally valuable to a large segment
of the industry. If such groups obtain broad support from within industry
the collective research group can overcome some of the ownership problems
of a private corporation investing in the research.

Property rights also influence the demand for products with particu-
lar characteristics. As discussed above in Section XII-B, without collec-
tive action the demand for public goods will be less than is socially opti-
mal. The national defense has long been recognized as a classic example
of a public good. The security of a country is a good which can be enjoyed
(consumed) by each resident of the country without excluding other residents
of the country. Thus, the social value will be greater than the value for
any individual or subgroup of individuals. This implies that the private
sector will have insufficient incentives to invest in national security.

The alternative nuclear fuel cycles have different characteristics
with respect to the risk of nuclear proliferation. Given that national de-
fense is a public good, the private sector will not recognize the full social
cost of proliferation risks. The incentives of the private sector are to
choose a production technology and mix of inputs which minimize the produc-
tion costs, but only the private costs. It is not surprising then to find
the utility industry relatively insensitive to the issue of nonproliferation.
If a reactor concept and fuel cycle offered lower costs to protect against

XII-10
potential military or terrorist threats but higher production costs, the
private sector would tend to recognize only the production costs. It is
then the proper role of government to intervene in the marketplace and
adopt decision criteria which accounts for the total social costs (both
private and extra-market) of a technology. There are a wide range of
institutional forms this might take.

Because of the scale at which most hybrid reactor designs produce
fissionable fuels, the Federal Government could maintain ownership of
the reactors and operate them as they have enrichment facilities in the
past. The ability to breed fuel for many conventional 1ight water reactors
of equivalent thermal capacity for each hybrid reactor makes them highly
amenable to centralized operation. If the Federal Government is the sole
user then some of the ownership problems of the research output are re-
duced. However, society could also lose the benefits of competition for
continualy eliminating inferior technologies.

The rights of entrepreneurs are often attentuated by special interest
laws and regulations which can influence incentives for innovation and
create commercialization barriers. For example, entry into the electric
generation industry is controlled by state regulatory commissions. Rigidi-
ty of regulators could inhibit hybrid commercialization as several charac-
teristics of hybrid reactors are likely to result in pressures for institu-
tional shifts among the users and the vendors of the technology. Because
hybrids are a joint production process with outputs of both the breeding
of fissile fuel generating electricity, there could conceivably be a sig-
nificant change in the industrial structure of the sector utilizing the
hybrid reactor. For example, as supplies of existing fuel become more
scarce and their finding costs more uncertain, then those firms engaged
in the traditional fuel supply would have strong incentives to enter the
market. Even if the steam from the hybrid reactor was not employed to
generate electricity, its operation is likely to be treated as a public
utility which, given the present regulatory framework, would raise a wide
range of legal and economic questions.

XII-11
There is some evidence of policy changes at the Federal level which
could encourage entry into the generation segment of the industry. In the
National Energy Policy Act now before Congress, there is a provision on
sales of electricity by non-utilities with cogeneration facilities. Of
course this is a small scale technology as compared to hybrid reactors;
however, it could be an important institutional shift.

E. CAPACITY TO PRODUCE

If the proper incentives can be secured through either private property
rights or collective action, then attention will turn to the decisions
of private firms to acquire the necessary capacity to produce. This element
in the process is a matter of information transfer or as it is sometimes
termed "technology transfer”. The entrepreneur must realize the technical
capability to offer a new product or process along with the potential for
economic gain through increased business activity or cost reductions.
Acquiring the capacity to produce is a focal point of the reduction of un-
certainties and risks.

Governmentally supported research has often used demonstration projects
as vehicles to transfer the technology and reduce at lTeast the technological
uncertainties. The use of demonstration projects can be an important tool
in the transfer of Federally sponsored research; however, they will not
in-and-of themselves guarantee success. A recent study revealed several (4
(1) The project should only begin after the principal technological problems
have been resolved; (2) Costs and risks should be shared with the private
sector or ultimate user of the technology; (3) Projects originating from
the private sector tend to have faster rates of diffusion than those initiated
by the Federal Government; (4) Faster rates of diffusion also occur where
there is an already existing market (buyers and sellers) for a related

factors which have been associated with successful demonstration projects:

product; (5) Successful demonstration projects tend to include all elements
necessary for full scale production and uses of the innovation; and

(6) Projects facing externally imposed time constraints were less likely

to be successful. Observations (2), (3), and (4) follow the general notion

XIT-12
that projects with significant early involvement of the private sector tend
to be more successful than those carried longer by the government alone.
This, of course, can be a chicken-and-egg argument. Good projects quickly
attract private interest versus private involvement will improve the

focus of the project. Irrespective of the direction of causality, the
private sectors interest can be an important signal to managers of publicly
sponsored R&D.

Some private sector analyses are expressing interest in the hybrid
reactor concepts especially for their breeder potential for LWR/HTGR fissile

fue]s.(]4)

However, it is also suggested that widespread involvement, especial-
1y financial involvement, may not be forthcoming. In the late 1980's, the
industrial vendors could potentially be heavily involved in the transfer to
HTGRs and faster breeder reactors. Also at this time, utilities will be
making substantial capital investments in additional generating capacity.
However, if the availability of fissile fuel becomes a significant problem

in the late 1980's and early 1990's, the utility sector will have strong
incentives to acquire the fuel breeding potential of hybrid reactors. The
adjustment costs to the private sector may be high (especially to vendors)

and, therefore, the direct costs of hybrid reactors will have to be more

than marginally lower to encourage diffusion.

Demonstration plants have been useful in reducing technological
risks and desseminating information. There still can be significant finan-
cial risks in the construction of the first few commercial scale plants.
In the case of fission reactors, the vendors offered purchasers "turn-key"
contracts which shifted the financial risks from the purchaser to the ven-
dor. The Federal Government has sometimes contributed financial support
to the first commercial units of a technology. Given that private indus-
try behaves in a risk adverse fashion protecting the purchaser from cost
overruns will be more of an incentive than an equivalent fixed subsidy. A
well designed policy would also require the vendor to share in the overrun
risk to insure cost effectiveness.

The decision of a private firm to acquire the necessary technical
capabilities and enter a new product market will interact with potential

XI1-13
variations in product characteristics. Firms which sequentially enter

a market will tend to offer a product with different attributes than
existing producers.(]s) This will afford them the greatest opportunity
for securing an economic profit. The General Atomics case with the HTGR
was discussed above, but it is again an interesting example. General
Atomic, without a prior market position in the electric generating plant
market, sought entry with a highly differentiated product from the exist-
ing producers terminal electric generators.

In Section XII-D control of fussion-fission technology by the Fed-
eral Government was considered as a means of installing the technology
into the economy. The focus in the present section is on the transfer of
technology to the private sector. In this respect hybrids concepts have
two factors influencing their adoption. First it appears as if the tech-
nological risks of hybrid reactors are less than for other advanced reac-
tor concepts. The fission portion of the reactor is well understood and
the fusion requirements are lessened by the energy multiplication of the
blanket. The second factor is the widespread appeal of hybrids for their
breeder characteristics. Each segment of the nuclear industry is affected
by the uncertainty surrounding fissionable fuels and thus have an incentive
to obtain the capacity to produce hybrid reactors. Current reactor ven-
dors may need to offer the technology in order to continue selling conven-
tion reactors. They will be limited internally by their ability to cope
with an additional technology. Fuel suppliers would also naturally find hybrids
attractive for extending the 1ife of the nuclear fuel business. Given that many
of the larger fuel supplies are horizontally integrated energy companies,
their activity in this area could be blocked by changes in anti-trust reg-
ulations. The third segment of the industry, the utilities given continued
growth in the conventional reactors, will be a strong incentive to adopt
the fusion-fission reactors, again for the capabilities to produce fissile
fuel. Which of the groups will act more aggressively in entering the
market will depend largely on the managerial costs of coping with the new
hybrid nuclear technology.

XII-14
F. PRODUCT CHARACTERISTICS

The attributes a technology offered the consuming public are the final
consideration in the commercialization process. This element is in many
respects the mirrow image of market demand. The ability of a new technology
to meet the requirements of the final users is the last hurdle in the com-
mercialization path. As was observed in the preceding section, the charac-
teristics offered in a market are in part determined by the structure of the
supply sector and the entry decision of competitors. Firms competing in a given
market will formulate strategies based upon several parameters, one of which
will be price; others will include service and specific product attributes.

In a recent analysis of successful corporate innovation policies,
Alan Fusfe]d(]G) introduced the notion of "technology demand elasticities".
This is an extention of the formula price elasticity concept which is an
index of the sensitivity of the quantity of a product demand to changes
in the product's characteristics. These elasticities will, of course,
vary from market to market. Fusfeld suggested seven generic technologi-
cal characteristics. They are as follows.

1) Functional Performance - basic task the device is

to perform
2) Acquisition Cost?- - capital cost of the device
3) Operating Costgf - variable cost per unit of service

4) Ease of Use Characteristics - "the form of the user's
interface with the device"

5) Reliability - normally required service and random
breakdowns

6) Serviceability - speed and cost of repair

7) Compatability - the ease in which the device can be
adopted into the existing system

a/ Perhaps it would be more meaningful to only consider the relative cost
of acquisition versus operating. This would be the cost of capital
to the firms compared to the real costs of variable inputs over the
1life of the device.

XI1I-15
The more important these elasticities are the more the market will be sub-
ject to extensive non-price competition. Each firm will endeavor to secure
some portion of the market which it can uniquely service and exercise a
degree of market power. If economies of scale in production are not sig-
nificantly relative to the market, then the probability of successfully
commercializing a new product is increased.

There have been a few studies on the characteristics of fusion reac-
tors from the point of view of the uti]ities.(]4’ 17, 18) However, little
attention has been given the hybrid concepts until recently. One comment
specifically toward the hybrid concepts suggests that development should be
continued very cautiously because it could associate the "fission related
difficulties and public-political animosity" with fusion reactors.(]4)
This is an interesting point. However, it is beyond the scope of this re-
port to evaluate and should possibly be addressed after the collection of

primary data on the public's reactions.

Characteristics of hybrid reactors appear to generally satisfy the
market's demands. Fusion-fission reactors are potentially the best breed-

ing alternative now under consideration. The hybrid concept has been shown
to be the most economical a1ter-nat1've.(]9 Also because of the number of
1ight water reactors each hybrid could support, siting requirements are

significantly reduced. Some driver design would have problems interfacing
with the electrical grid. This is the objective of future research. The
scale of fusion reactors in general has sources of criticism.(]7’ 18) This
too can be addressed in future research. However, it should be pointed out
that institutional structures are continuously being altered due to tech-
nological change. System growth will accommodate increases in plant scale,
as will improvements in transmission technology. In addition, individual
utilities will become more comfortable with recent organizational inno-
vations lowering their transaction costs of involvement in system inter-
ties and regional power pools. MWith respect to proliferation resistance
fusion-fission reactors can be compatible with any previously selected

fuel cycles. Also hybrids have the potential isotopic tailoring to reduce
proliferation risks.

XII-16
The acceptability of a technology's characteristics in meeting the
demands of the marketplace is the ultimate test in the commercialization
process. Numerous analyses of the problems of technological change sug-
gest that significant user input early in the development can encourage
the match of capabilities and needs. It is also advantageous to remove
burdensome regulatory and institutional barriers which often do little

more than protect special interest groups. Also, free entry into the
new industry should be encouraged to the fullaest extent possible.

G. CONCLUSIONS

Before summarizing the findings presented in this chapter, it is impor-
tant to consider the fundamental problems of managing R&D investments in
the public sector. The principal argument for governmental intervention
in civilian technological change, positive externalities, is very difficult
to apply in a general decision rule. The notion of spillover benefits can
be attractive politically, but unfortunately it can be misused to justify
programs which simply fail to have sufficient social and private returns
to justify the investment. The inability to capture all the returns from
an investment is not the only distortion affecting the private R&D market.
The market structure and nature of the basic product may be such that com-
petition is channelled into non-price area, product differentiation. This
can lead to a significant amount of R&D investment for the firm to maintain
its market share. This may or may not be socially beneficial. If it is
not, there will be a tendency for over-investment in R&D to improve the
fim's products. The market is subject to further distortions due to
other policies of the government; this includes environmental and product
regulations, procurement practices, anti-trust, patent and copyright laws
and tax laws. Within this environment it is difficult to determine if
there are insufficient incentives for the private sector to invest in R&D
because of the externalities.

The existence of externalities from investments in R&D tend to cre-
ate additional problems for the management of government sponsored research.

XII1-17
There is a general tendency to model government research management sys-
tems after those of the private sector.(zo) Any decision process will be
tied to the nature of the incentive system. Thus, it may be difficult if
not self-defeating to make government R&D sponsors behave as a private
firm. The effectiveness of private decision making is linked to the

residual claims on the return to the firm.(21)

It will be impossible to
replicate this in the public sector. Further, it is difficult to predict
and measure the external benefits from a particular innovation. There-
fore, modeling public decision making after the private sector's will
generate a similar bias against those projects which produce the most ex-
ternal benefits. This is not to suggest that the public sector be

immune from the basic resource allocation rules, but rather to point out

the fundamental dilemma in managing the production of public goods.
The demand for hybrid reactors appears to be fairly straightforward

as a stepping stone to pure fusion and as a breeder of fissile fuel, that
is provided that can be cost effective. The hybrid nonproliferation attri-
butes do represent a "public good" with their inherent problems. Obtain-
ing desirable operating characteristics in terms of reliability and com-
patability will require concerted design efforts and practical input from
the users. The transfer of the technology and entry decision by firms into
the market will be very significant. If a fusion-fission reactor concept
becomes technically successful, it will potentially imply some very inter-
esting structural changes on both sides of the market--venders and users.

XII-18
H.

10.

11.

12.

13.

14.

SECTION XII REFERENCES

W. J. Abernathy and J. M. Utterback, "Patterns of Industrial Innovation,"
Technology Review 80(7):40, June/July 1978.

Rate and Direction of Inventive Activity, National Bureau of Economic
Research, Princeton University Press, Princeton, NJ, 1962.

M. I. Kamien and N. L. Schwartz, "Market Structure and Innovation: A
Survey," Journal of Economic Literature. 9(1):1, March 1975.

A. B. Linhares, An Overview of Federal Technology Transfer NTIS PB-255,
U.S. Department of Transportation, June 1976.

F. M. Sherer, Industrial Market Structure and Economic Performance. Rand
McNally, Chicago, IL, 1970.

Supply 77, Electric Power Research Institute, Palo Alto, CA, May 1978.

Charles River Associates, Uranium Price Formation, EPRI-EA-498, Prepared
for the Electric Power Research Institute, Palo Alto, CA, October 1977.

"CONAES Waxes Gloomy on Uranium," The Energy Daily, Washington, D.C.,
p.2-3, August 7, 1978.

Government Support for the Commercialization of New Energy Technologies,
Massachusetts Institute of Technology, MIT-EL-76-009, November 1976.

J. Demsetz, "Some Aspects of Property Rights," Journal of Law and
Economics, 9(3):61, October 1966.

"Toward a Theory of Property Rights," American Economic Review,
LVII(3):76, May 1967.

"DOE Reserves Stance, Grants Gulf Foreign Patent Fights for SRC-II,"
Inside DOE, New York, page 2, June 26, 1978.

W. S. Baer, L. L. Johnson, and E. W. Merrow, Analysis of Federally Funded
Demonstration Projects. R-1926-DOC, Rand Corporation, Santa Monica, CA,
April 1976.

M. Lotker, "Commercializing Fusion," presented at American Nuclear
Society/European Nuclear Society, International Meeting, Washington,
D.C., November 1975.

XII-19
15.

16.

17.

18.

19.

20.

21.

E. C. Prescott and M. Visscher, "Sequential Location Among Firms with
Foresight," Bell Journal of Economics, 8(2):378, Autumn 1977.

A. R. Fusfeld, "How To Put Technology into Corporate Planning,”
Technology Review, 80(6):51, May 1978.

C. P. Ashworth, A User's Perspective on Fusion, Pacific Gas and Electric
Co., presented at the annual Conference of the American Association for
the Advancement of Science, Denver, CO, February 1977.

W. C. Wolkenhauer, "Interface of the Controlled Nuclear Fusion Program
with Utilities," Mimeo, Washington Public Power Supply System, 1977.

D. J. Dreyfuss, B. W. Augenstein, W. E. Mooz, and P. A. Sher, An
Examination of Alternative Nuclear Breeding Methods, R-2267-DOE, Rand

Corporation, Santa Monica, CA, July 1978.

G. Eads, "U.S. Government Support for Civilian Technology: Economic
Theory vs. Political Practice," Research Policy 3:2, November 3, 1974.

R. T. Masson, "Executive Motivations, Earnings, and Consequent Equity
Performance,” Journal of Political Economy, 79(6):1278, December 1971.

XII-20
XIII. TECHNOLOGY STATUS AND RD&D REQUIREMENTS
A. PRESENT STATUS OF FUSION PHYSICS

1. Tokamak

Tokamak fusion research in the U.S. is being conducted at a number of
national laboratories and universities. The major ongoing experiments are
located at Princeton Plasma Physics Laboratory (PPPL), Oak Ridge National
Laboratory (ORNL), General Atomic Corporation (GA) and Massachusetts Insti-
tute of Technology (MIT). Research directions at these laboratories are
summarized in Table XIII-A-1. A list of current or planned U.S. Tokamak
experiments are tabulated in Table XIII-A-2.

TABLE XIII-A-1. U.S. Tokamak Research

Laboratory Research Direction

PPPL Demonstrate Scientific Feasibility of
tokamak fusion, evaluate divertor per-
formance and supplementary heating tech-

niques.

ORNL Examine efforts and means of reducing plasma
impurities developed from plasmal wall inter-
actions.

GA Evaluate stability and performance of doublet
cross section tokamaks.

MIT Explore plasma confinement in high magnetic
fields.

The principal measures of progress in tokamak fusion physics are the ijon
temperature Ti’ plasma density n and energy confinement time t. The product
of the last two nt has been termed the Lawson number. For tokamaks operating
with ion temperatures near 10 keV, the Lawson number must exceed about 10]4
in order that energy losses from the plasma are balanced by fusion energy.

s/cm3

Figure XIII-A-1 shows recent and expected progress in achieving high temper-
atures and nt in both tokamak and mirror experiments.

Experimental results from the currently operating devices give encour-
aging signs for the success of the large, two-component tokamak TFTR under
construction at PPPL. PLT has shown an nt product of 1013 s/cm3 with as high

XIII-1
Experiment

PLT

ORMAK

ISX
Microtor
Macrotor
Doublet IIA
Alcator A
Alcator C
PDX

Doublet III
TFTR (2)

TABLE XIII-A-2.

R{cm)

130
80
92
30
90
66
54
64

140

140

265

U.S. Tokamak Experiments

a(cm)

45
26
26
10
45

30/100

9.5
17
45
45

110

(a) To begin operation in 1981

XITI-2

Btor(T)

4.2
2.5
1.8
6/25
2/7
0.8

10

14
2.4
2.6
5.6

Laboratory

PPPL
ORNL
ORNL
UCLA
UCLA
GA
MIT
MIT
PPPL
GA

PPPL
CONFINEMENT MEASURE

nr (cM-3 sEc)

1015

R
ARRIE 1
ADIAT! 2 0 0 FUSION
POWER
REACTORS IGNITION
10 = GAIN

1980 \ ! 1982 1
“\D\'_'“ ) ‘I.FTE,\ /
" v— .’_.\
1013
EARLY
TOKAMAKS
101 103
MIRRORS fzfi)
] A 1975 h

0.1 1.0 10 100 keV

TEMPERATURE (keV)

FIGURE XIII-A-1. Technical Progress and Outlook

in Magnetic Fusion

XIII-3
as 6 keV temperatures, while ISX at a lower magnetic field has demonstrated

a plasma beta value (ratio of plasma pressure to confining magnetic pressure)
of 6%. The MIT Alcator A experiment, at considerably higher density and
magnetic field, has demonstrated nt in excess of 10]3 s/cm3 at a temperature
of 1 keV. In this case the effective charge of the plasma is unity (no
impurities), while in PLT it is a less desirable Zeff = 2. The Alcator C

14 s/cm3, or near

and Doublet III experiments are expected to achieve nt > 10
the Lawson condition for reactor ignition.

Tokamak theory has been developed to the point where many experimental
results are well explained. This is particularly true for macroscopic plasma
performance. Mechanisms of energy loss from the plasma are not fully explained
however, and observed electron heat conduction losses are larger than the
- neoclassical prediction by a factor of 10-500. The theoretical uncertainties
in predicting energy loss at conditions near those required for fusion reactors
have prompted physicists to establish empirical scaling laws which relate
energy confinement times to plasma parameters such as density, temperature
and size. Fortunately, the data base to do this is strong and results from
a variety of confirming diagnostic techniques. A recent assessment(z) of the
tokamak confinement data base notes it to be basically sound and credible.

The data has been satisfactorily determined by an acceptable computerized
compilation of many diagnostic techniques, in numerous laboratories, with
results showing an impressive consistency. The new and planned experiments
of the DOE-OFE tokamak confinement program is expected to reinforce or help
establish the data base in areas of auxiliary heating, impurity control,

B limits and elongated plasma.

2. Mirror

The major mirror fusion research is being conducted at the Lawrence
Livermore Laboratory (LLL). Other laboratories which have mirror programs
include the University of Wisconsin and Cornell University.

At LLL three mirror devices are operating or under construction: The
Beta I (formerly 2XIIB) the Tandem Mirror Experiment (TMX) and the Mirror
Fusion Test Facility (MFTF).

e Beta II relies on magnetic fields to confine a hot, dense
plasma for a short time. It features C-shaped magnetic coils

XIII-4
that form the confining magnetic field. Their unique shape (in
what is known as a yin-yang geometry) stabilizes the confined
plasma by creating a magnetic field (a magnetic well) that in-
creases in every direction from the plasma center.

® MFTF, now being constructed, will bridge the physics and engi-
neering gaps between current experiments and an experimental
fusion reactor planned for operation by 1990. MFTF will use a
superconducting magnet of yin-yang design (similar to the 2XIIB
experiment). This magnet will be capable of continous operation.

e The tandem mirror reactor concept consists of a long solenocidal
magnet terminated at both ends by conventional mirror cells.
These cells will act as "end plugs" to prevent plasma leakage
out the ends of the solenoid. TMX is being constructed to
test principles of this concept.

Experiments with 2XIIB have shown that startup can be done in steady-
state magnetic fields and that scaling of the density n confinement time =

product follows the classical relationship nt ~ w13/2

up to a mean ion

energy Ni = 13 keV for injected'powers up to 3/MW at 20 keV. This device
also demonstrated operation with g = 2.5. It implies a close approach

to a fie]d-revefséd state. In a field reversed mirror p1asma, a'ring-

shaped plasma between mirrors is formed of sufficient density to create

a locally field-reversed region by virtue of its ion diamagnetic currents.
This would significantly augment the plasma confinement of the mirror machine

and thereby enhance its Q.

On the basis of the favorable plasma physics results with 2XIIB, a
larger experiment, the MFTF is being constructed. It is scheduled for com-
pletion in Tate CY 1981. MFTF will test further scaling of mirror plasma
confinement and will investigate advanced engineering problems such as those
associated with NbTi superconducting magnets, neutral beam injectors, plasma
wall interactions, disposal of neutral particles and ions escaping from the
plasma chamber and high speed vacuum pumping techniques.

The TMX will test a new principle for improved confinement in mirror
systems. The basic idea is to reduce the plasma loss rate by electrostat-
ically plugging the ends of a solenoidal central confinement region using

XIII-5
the high positive ambipolar potential generated in minimum-B end plugs. Each
end plug will be driven by the injection of neutral beams from 12 source
modules, in a manner similar to that used in the 2XIIB experiment.

TMX has three fundamental objectives:

e To demonstrate the establishment and maintenance of a potential
well between two mirror plasmas.

e To develop a scalable magnetic geometry, while keeping macroscopic
stability at high beta.

e To investigate the microstability of the plug-solenoid combination
to maximize the plug-density/injection power ratio. Possible
reactor implications include the study of enhanced radial trans-
port in the solenoidal cell and the accumulation of thermalized
alpha particles in the central plasma. The TMX is currently in
the initial operation of "shake down" phase.

The projected mirror hybrid represents about a four-fold increase in
size over the MFTF, and the hybrid Q value is about 10 times that expected
for MFTF. Given continued progress with mirror-stability physics, the
mirror hybrid is a genuine near-term possibility, even though it has an
uncomfortably large recirculating power fraction. It should be noted that
the LLL-GA hybrid desingers have assumed attainable positive-ion neutral-
beam technology and NbTi superconducting technology in their hybrid design.

It is important to note that present classical end-1oss scaling behavior
in mirrors is obtained by injecting cold plasma or neutral gas at the ends.
This causes a heat loss, leading to Tow electron temperatures and low ambi-
polar plasma potential. Further physics research is needed to remove these
effects while retaining stability. It is expected that MFTF will demonstrate
whether or not the stabilizing cold gas or plasma can be dispensed with.

Phaedrus is a tandem mirror device (similar in design to TMX but smaller)
which is in operation at the University of Wisconsin. It will be used to
develop RF heating for the TMX. If the RF heating experiments are success-
ful, this technique could lead to a large reduction in the neutral-beam
heating required for a tandem mirror reactor and would significantly decrease

XITI-6
technology requirements and costs. Phaedrus will also be used to explore the
trapping of plasmas (i.e., reactor refueling) by RF techniques.

3. Linear Theta Pinch

The primary advantages of the linear theta pinch are its simple magnetic
configuration, known heating and ease of access to the core as a reactor.
Plasma heating to thermonuclear temperatures in the 4-10 keV range is under-
stood and practicable. There are no serious stability problems or problems
of confinement of the plasma across its magnetic field, which has simple,
longitudinal straight lines. The central problem is that of confining the
plasma along its length. However, recent experiments and theory show that
material end plugs successfully stop plasma particle flow. The remaining
energy-loss mechanism is that of thermal conduction by electrons and ions
along the magnetic lines to the end plugs. The energy loss time by thermal
conduction is sufficiently large to sustain the reactor energy balance and
to provide fissile production. There is a gross instability of linear theta
pinches wherein plasma rotation produces a wobble of the plasma column.
However, it does not lead to wall contact. In Scylla IV-P and STP this mode
is stabilized by magnetic line tying and in the latter case by wall stabili-
zation.

In 1964 the 1-meter Scylla IV produced an ion temperature of 5 keV at an
nt value of 5 x 10]0 cm'3 sec. A successful test of the staging principle,
on which reactor designs are based, was made in 1976 when the 4.5 m LASL
Staged Theta Pinch (STP) produced 2-keV plasmas using separate shock heating
and compression sources with adjustable plasma compression. A third important
test of the linear theta pinch has been the solid-end-plug experiments on the
5-m LASL Scylla IV-P device. Application of LiD plugs results in stopping
the flow of plasma particles. Thermal conduction at the ends of theta pinches
was tested in 1965 in Scylla IV and 1978 in Scylla IV-P and found to agree

with the theoretical predictions.

4. Inertial Confinement

The intertial confinement program is advancing with an array of short
pulsed energy drivers. These include lasers, 1light paricle beams (electrons
and ions) and heavy particle beams. |

XIII-7
a. Lasers

Lasers were the only ICF driver candidates given serious consideration
in the U.S. before 1972. 1In the early part of the ICF program, major
development efforts were established for high energy, short pulse laser
systems using solid (neodymium:glass) and gaseous (002) media.

Neodymium:glass laser technology is the most highly developed short-
pulse, high-peak-power laser technology existing today. The major ND:
glass systems development and target experiments are centered at Lawrence
Livermore Laboratory. Research experiments in the laser-plasma interaction
area are also being conducted using ND:glass lasers at KMS Fusion, Inc.,
the University of Rochester, and Naval Research Laboratory.

The ARGUS laser, which began operations at Lawrence Livermore Labora-
tories at the 2-4 TW level in late FY-1976 served as a prototype for the
20-beam SHIVA system. The first full power fusion experiment with the
SHIVA laser system took place in May, 1978. The 20-Arm SHIVA system
focused 26 TW of optical power on a deuterium fuel pellet yielding
7.5 x 100
significant thermonuciear burn where the fusion energy produced is several

14 MeV neutrons. Later experiments are expected to demonstrate

percent of the laser energy delivered to the target. At this time, glass
lasers are not viewed as a candidate for ultimate commercial fusion
applications because of probable pulse rate and efficiency limitations.
These lasers are being developed however for intermediate programmatic
tasks.

Short-pulse C0, lasers are being developed at Los Alamos Scientific
Laboratory (LASL) as drivers for laser fusion experiments. The resulting
gas laser technology, particularly the high efficiency (to perhaps 10%),
is considered extrapolable to repetitively pulsed laser designs which
will be required for the development of commerical fusion drivers. A
2-beam CO2 system has operated at 0.8 TW on each beam at LASL and has
produced neutron yield in early 1977. It is a prototype for an 8-beam,
10-20 TW system which consists of four of the 2-beam modules. This
8-beam system was successfully fired in mid-April, 1978 at the 8.4 kJ
output energy (15 TW output power) level. The beams will later be
fired at fuel pellets to initiate fusion reactions. Its goals are to

XITI-8
develop targets for ANTARES, study thermonuclear burn scaling, and to demon-
strate 20 times liquid density compression.

The projected near term achievements of the glass and 002 laser programs
are provided in Table XIII-A-3.

TABLE XIII-A-3. Office of Laser Fusion Physics Through Mid 1980s

Scheduled
Completion Date Anticipated
Designation Lab Power, TW Laser Drivers Results
13
Shiva (Nd:glass) LLL 20-30 Operational 101 N/Pulse
0
(10  N/Pulse achieved)
Nova - (Nd:glass) LLL
- Phase I 100 1982 Gain of 1 or more
- Phase I1I 300 1984 Gain of 20 to 100
Eight-Beam System (CO,) LASL  10-20 1978 10" °-102 N/Pulse
Antares (C02) LASL  100-200 1982 Gain of 1 to 8

b. Light Particle Beams

Light particle beam accelerators have been candidate drivers for inertial
confinement fusion since 1972. The early accelerators have been developed for
use in weapon's effect simulation studies. Since that time, much of the ICF
light particle beam program has been centered at Sandia Laboratories,
Albuguerque (SLA) and has been working to extend the useful range of machine
operation. Additional light particle beam work is being carried out at the
Naval Research Laboratory, Cornell University, Maxwell, and Physics Inter-
national. The significant new operating requirements for an ICF driver are
short pulses (10-30 ns), high instantaneous power (30-100 TW), a high energy/
pulse (1-10 MJ), good beam focus (1-5 mm dia.), remote beam delivery (1-5m),
repetitively and pulsed operation (1-10 Hz). |

Sandia's primary accelerator development tasks have involved electron
beam machines. Electron beam experiments with D-T targets using the Proto I
accelerator (2 TW, 400 KA, 3 MeV, 24 ns pulse) have produced neutron yeilds
greater than 106. Proto II (8 TW) has been in operation since 1977 and is

expected to yield additional beam-target coupling data in 1979. Scientific

XIII-9
breakeven (pellet thermonuclear output equal to beam energy input), is expected
with EBFA-I (30 TW) or EBFA-II (60 TW) by the end of 1985. In the development
of the latter two facilities it is felt that most of the driver technology
problems for commercial 1ight particle driven inertial confinement fusion will
be solved, except for those dealing with long-life rep rate, and operation in
nuclear environment. Work is also underway to develop a 10 Hz, 10 KJ machine
during the next five years to address problems associated with repetitive
operation.

Recent developments indicate that these electron accelerators can be
converted to ion accelerators with relatively minor changes. Work is in
progress to access the beam generation efficiency of such a converted system.
Success would allow the use of light ions (carbon and 1lighter) and reduce
problems associated with electron-beam preheating of the pellet fuel material
caused by electron penetration. Satisfactory light ion operation must be
demonstrated before the impact of this option can be assessed. EBFA-I and II
have been designed to operate with either polarity to allow modification should
a light ion diode be developed successfully.

c. Heavy Ion Beams

Heavy ion beam driver systems under consideration at this time are based
on the accelerator development and operating experience gained from high
energy physics experiments. This experience spans a period of more than 40
years and includes participation by Fermi Lab., Brookhaven National Laboratory,
and Lawrence Berkeley Laboratory, each having different but complimenting
accelerator technology.

Comparing the present capability to the anticipated needs for a success-
ful heavy ion fusion driver reveals that (1) the particle energies achieved
in recent physics machines greatly exceed the needs for a fusion driver,
(2) the technology must be demonstrated for heavy ions and very large instan-
taneous current levels (Existing machines already have demonstrated large
energy per pulse, 4 MJ 15% and large average power levels, 0.54 Mw 85%.)
and (3) the existing accelerator technology also has a demonstrated capa-
bility for pulsed operation that exceeds the rate anticipated for inertial
fusion applications. This pulsing capability is very significant for commercial

XITI-10
and programmatic needs even though it is not useful for the near term for proof
of scientific feasibility.

The existing physics machines have established performance records that

document their ability to provide:

- good operating efficiencies
(overall systems: up to 15%)
(subsystems: up to 42%)
- good machine reliability and availability
(ZGS at Argonne National Laboratory/85% of scheduled time)

The efficiency values are indicative of attainable values but do not
represent an upper limit since this has not been emphasized in past research.
Improvements can be anticipated with increased emphasis on this problem.

d. Fusion Targets

At the present stage of inertial-confinement, fusion neutron yields in
the range of 109 - 1010 per shot have been obtained. The targets have gener-
ally been thin-shelled glass or metal submilimeter microspheres containing
D-T gas at several hundred atmospheres. These targets explode as a reaction
to the laser, e-beam or ion-beam - initiated surface vaporization. Compression
heating to thermonuclear conditions occurs as a result of the initial impulse
applied to the surface; this is termed an exploding "pusher" target. For
practicable hybrid drivers multilayered high-gain targets must be used.(3)
Whether such targets can be fabricated for economical commercial application
will require extensive research and development.

XITI-11
B. FUSION DRIVER RD&D REQUIREMENTS

1. Tokamak

The technological development of the ignited tokamak fusion driver of the
Tokamak Hybrid Reactor (THR) can fit into the DOE-OFE confinement and D&T
programmatic schedule which plans to have operational a pure fusion device
about the year 2015 (Figure XIII-B-1). The progress of the major facilities
are listed in Table XIII-B-1. The majok plasma physics input for an ignited
Tokamak Engineering Test Facility (TETF), which could conceivably be an
Hybrid Experiment Facility (HEF), would come from the U.S. devices through
TFTR in addition to ORMAK-Upgrade, Alcator C, et al., as well as foreign
experiments (Table XIII-B-2). The D&T requirements (Figure XIII-B-2 and
Table XIII-B-3) would come from the beam development for TFTR, the Large Coil
Project (LCP) for the superconducting magnets, the High Intense Neutron Facility
(HINF), the Multi-Component Radiation Facility (MCRF), and Fusion Materials
Irradiation Test (FMIT) for the materials qualifications, and the Tritium
Systems Test Assembly (TSTA). The HEF features would allow its operation as
early as c. 1989. The features which may represent some question include
the matter of stabilizing a "D" shaped MHD equilibrium plasma which has already
been demonstrated on several tokamaks (e.g., Versator, Rector, T0-1) but would
benefit from even further study. The technological development of divertor
collection systems should perhaps be more clearly defined in the D&T program.

The successful operation of the HEF would impact the final design and con-
struction of the scheduled EPR which with a hybrid blanket and appropriate fuel
and blanket remote handling capabilities could conceivably be an Hybrid Experi-
mental Reactor for operation c. 1994. Such a facility would produce power and
demonstrate an integrated tritium handling and refueling capability. Its
operation in turn would impact the final design and construction of the mag-
netic fusion DEMO which could be a Prototype Hybrid Reactor (PHR) to operate
c. 2000 that would precede the first Commercial Hybrid Reactor to be built and
operated early in the next century.

XIII-13
vL-11IX

FiSCAL YEAR 78 {79 81|82 83{84 (8 |8 {87 |88 189 90|91 192 193194195 |961}97 {98 2@5 2010 2015
DPR (OR PHR) L—L l
EPR (OREHR) fi? —

ETF (OR HEF) ¢ 9
TOKAMAK ETF
MIRROR ETF _!
ALTERNATE ETF
TFIR
MFTF * PY
ALTERNATE SCALING “_*__.# BEGIN OPERATION
EXPERIMENTS BEGINTITLE 1
PDX DIVISION DECISION
D OCCURS 1 YEAR EARLIER
T™X L o+-o BEGIN CONCEPTUAL DESIGN
ALTERNATE PROOF [NITIATE STATEMENT
OF PRINCIPLE TESTS COMPLETE STATEMENT

FIGURE XIII-B-1.

Major Facilities Schedule

TABLE XITI-B-1.

Reactor

TETR (Driven
Tokamak Engineer-
Test Reactor

EPR
Experimental Power
Reactor

DPR

Immediate
Supporting Year of
Devices Operation

D-III, PDX, PLT, 1990-92
TTA, RTNS, TFTR,
JET

TETR, TFTR, JET, 2000-04
T-20-JT-60

TETR, EPR 2010-15

XIII-15

Objectives of Major Fusion Reactor Facilities

Objective

Test Materials to 102!

n/cm?2

Fueling (E = 14 MeV)

S/C Magnets

Limited T Breeding

Neutronics Test

Remote Handling

Blanket Design Tests

Performance Test of Plasma
Operation Required for
Hybrid

Limited Electrical Gener-

ation

High Temperature Operation

Fabrication of CTR Vessel
Components in Field

Structural Material Test
(Fatique)

Demonstrate Safe Handling
and Pumping of Liquid
Metals in CTR Environment

Reliability of S/C Magnets

Remote Assembly and Dis-
assembly

Breeding and Containing
Jritium

Demonstrate Safe Reliable
Power Generation in a
Reactor System which
Scales Readily to a
Commercial Reactor
TABLE XIII-B-2. Features of the Tokamak Fusion Driver(]3)

Related to Large Tokamak Experience

Relevant Preceding(a)
Feature Implication Experiments
High Neutron Flux TCT Operation TFTR, JET
and Small Size
Noncircular Plasma D-III, JET PDX-UG
High t_ and Low Zeff Divertor, gas blanket DITE, JFT-2a/DIVA
P PDX, ASDEX, ISX

Long Burn Time Pellet Fueling ORMAK, ISX
( ~30-60 s)
Long Pulses at Superconducting Toroidal T-7, T-10M
Reasonable Power Field Coils Large Coil Project
Costs MFTF
High Power Neutral Efficiency ~ 50% at Beam Test Stands
Beams 150 keV
(a) TFTR = Tokamak Fusion Test Reactor (PPPL)

JET = Joint Eurcpean Torus (EEC)

D-IT11I = Doublet-III (GA)

PDX-UG = Poloidal Divertor Experiment-Upgrade (PPPL)

DITE = Divertor Injected Tokamak Experiment (Culham)

JFT/2a = Japanese Tokamak with Divertor (JAERI)

PDX = Poloidal Divertor Experiment (PPPL)

ASDEX = Axisymmetric Divertor Experiment (MPI-Garching)

ORMAK = Oak Ridge Tokamak (ORNL)

T-7, T-10M = Tokamak-7, 10 Modified (Kurchatov)

MFTF = Mirror Fusion Test Facility (LLL)

XITI-16
LL-ITIX

FISCAL YEAR 78 179 | 80|81 (82 {83 {84 |85 |86 |87 (88 |89 |90 |01
BLANKET AND SHIELD -~ +- ®
TSTA °
FMIT <
MCRF A——o+0

® BEGIN OPERATION
HINF —e ® BEGINTITLE L
DIVISION DECISION

LARGE COIL PROJECT @ o OCCURS 1 YEAR EARLIER

HIGH FIELD TEST FACILITY |

——9

A BEGIN CONCEPTUAL DESIGN

FIGURE XIII-B-2. Engineering Facilities Schedule

TABLE XIII-B-3. Objectives of Major Fusion Engineering Facilities

Year of

Facility Operation
Blanket and Shield Facility 1988
Tritium Systems Test Assembly (TSTA) 1982
Neutron Source Facilities

(FMIT, MCRF, HINF) 1979-83
Large Coil Project 1982
High Field Test Facility 1980

XIII-18

Objective

Test prototype blanket
and first wall structures.

Test thermal/hydraulic
performance and electro-
magnetic compatibility.

Test capability to accomo-
date accident conditions.

Demonstrate vacuum integrity
and remote maintenance
operations.

Test neutronic models and
performance.

Demonstrate safe and
economic handling of tritium.

Test small material samples
in fusion neutron environ-
ments.

Test large superconducting
magnet designs.

Test high field supercon-
ducting magnet materials.
2. Mirror

It is generally recognized that, even with the energy multiplication of
a fissile blanket, the low Q value of the classical mirror gives it an
excessive recirculating power fraction and therefore poor economic performance.
In addition, the open ends in near spherical geometry, as well as the beam
injection parts, lead to poor blanket coverage of the plasma neutrons. For
these reasons the mirror fusion program has been redirected to the tandem-
mirror confinement concept for reactor applications.

A central feature of a mirror fusion device is the positive ambipolar
electrostatic potential which the plasma assumes to keep the electrons from
escaping faster than the ions. This positive potential is made the basis of
a new end-plugging method for a linear solenoid by using two minimum-B
mirrors at the solenoid ends to contain its jons electrostatically along the
axis. The plugs are high density mirror devices (plasma volume Vp) whose (
values are less than unity. However the Q value of the composite system
with central linear plasma volume VC can be raised to larger values by choos-
ing VC/Vp large enough. As a test of these principles, the Tandem-Mirror
Experiment (TMX) is now in operation at Lawrence Livermore Laboratory (LLL).

A pure-fusion system is envisaged as a 650-MWe system with 1-MeV neutral-
beam injection into the end plugs, a first-wall radius of about 1 m and a
length of about 80 m. The maximum plug magnetic field is 16.5 T and the
central-cell magnetic field is 2.2 T.

As of September 1978, no detailed description of a tandem-mirror hybrid
has been published by LLL. However, an outline of such a design based on the
arrangement of Figure XIII-B-3 has been made. Like other linear devices it
has the advantage of simple modular construction and it can be attractively
short with small power rating.

A major RD&D requirement for a Mirror Hybrid is to run a tandem-mirror
experiment and to check the main new features of its operation. The end-plug
physics is like that of the existing 2XII-B or the projected MFTF. However,
there are substantial questions of the stability of the new geometry, which
includes regions of bad magnetic curvature. The thermal conduction problem
may be aggravated owing to the very high density of the end plugs and attendant
high plasma energy flux on the end walls.

XITI-19
Technological questions include the radiation-hardened injectors and the
superconducting magnets of the central solenoid, as well as the high-field
end plugs. Such development would follow successful operation of TMX and
MFTF as scheduled in Figure XIII-B-1. The selection of the tandem-mirror
concept for the Engineering Test Facility would also be based upon their
operation. As with the tokamak hybrid, the Tandem-Mirror Engineering Test
Facility could conceivably be a Hybrid Experimental Facility whose blanket
modules would be a test bed for hybrid blanket and fuel development. The
other scheduled magnetic fusion engineering development facilities of Figure
XIII-B-2 would also fulfill the D-T requirements for the Tandem-Mirror
hybrid development in support of the scheduled operation of the EHR and PHR.

Helium Circulator
Steam Generator \
<a»
Central Cell -

Injection

Cryopumps

Plug Coil

Plug
Injection

Blanket Service Area

End Leakage Tank
(Direct Converter)

FIGURE XIII-B-3. Tandem Mirror Reactor

XIII-20
3. Linear Theta-Pinch

The main technological developmental requirements for the Linear Theta
Pinch Hybrid Reactor (LTPHR) include the development of pulsed electrical
energy storage. Conceptual designs based on experience with superconducting
machinery indicates that homopolar motor generators would furnish the plasma
compression power economically and at sufficiently high efficiency. A
development program for homopolar energy storage should therefore be pursued
as part of the engineering development facilities schedule of Figure XIII-B-2.
Since proof of principle tests and design studies have indicated that the
first-wall pulsed heat loads and thermal stresses of the LTPHR can success-
fully be withstood, larger-scale simulation tests should also be included
in the development schedule. The relatively small-diameter superconducting
magnets required for LTPHR are little beyond the current state of the art and
could be developed in the near term as part of the magnet development program.

Regarding the major reactor facilities for the development of the LTPHR,
on the basis of present experimental results a scientific feasibility demon-
stration could be performed in a stage physics experiment of approximately
100 meters in length. It would have pulsed energy storage prototypical of
the LTPHR but on a shorter time scale. As with the TFTR, the plasma would
initially be deuterium, followed by D-T. It is already too late to schedule
such a facility for selection of the ETF fusion driver as indicated by the
magnetic fusion facility schedule of XIII-B-1. However, its successful
operation (c. 1994) could impact the driver for the Experimental Power Reactor
(EPR) (c. 2004) or the Experimental Hybrid Reactor (EHR) preceding the PHR.
Such a facility would entail a 500 meter LTP with full hybrid blanket coverage
and all of the subsystem engineering features the commercial LTPHR would have.

4, Inertial Confinement

In order to develop an inertial confinement hybrid (ICH) it will be neces-
sary to conclude the physics research (driver/pellet interaction studies) and
proceed through engineering development to a prototype hybrid reactor. In
addition to lasers, R&D must be carried out for light-particle beams, heavy
ion beams, and fusion targets.

XITI-21
a. Lasers

The precise requirements for the ICH laser cannot now be specified. The
conceptual requirements provided in the LLL/Bechtel design are:

Laser Energy (Selenium laser with 489 nm wavelength) 400 KJ
Laser Efficiency 1.2-1.5%
Laser Pulse Repitition Rate 5-8 Hz
Pellet 250

This laser was selected because it is known to have a short wavelength which is
currently preferred for pellet coupling. It is one of several candidates dis-
cussed in Section XIII A.4 above.

None of the high power lasers available today satisfy all the criteria
for a commercial driver (reliable, efficient, pulse rate capability, good
pellet coupling, peak pulse power, and pulse shape). As a result, the
identification and characterization of new laser media and excitation tech-
niques are being carried out to develop an advanced laser capable of driving
a fusion reactor. This is being done by:

e Conducting a program of fundamental research to identify new laser
candidates,

e Evaluating the usefulness of present advanced laser candidates for
commercial fusion application,

e Initiating an effort to scale an efficient visible laser to the 1 KJ
level (1 nsec pulse width), and

e Conducting a program of supporting technology to aid in scaling high
power lasers.

e The primary candidate advanced lasers currently being evaluated are:

e Rare earth molecular vapor (A = 0.545um) - terbium - aluminum - chloride
comples.

e HF. chemical laser (x = 1.315u) 0 a high gain medium pumped by
chemical reactions (expected efficiency is 5%)

XIT1-22
e Iodine (A = 1.315um) - a low gain medium in the near infrared spectra
whose efficiency needs improvement (expected efficiency is 1-2%)

e Metal vapor excimer lasers (i = 0.173um-0.485um).

The initial laser chosen for the power scaling experiment (oxygen, X =
0.557um) is from the Group IV metastable atom lasers.

The present advanced laser development plan includes:

e Selection of several candidate lasers during FY-80 for development of
1 kd modules

e Completion and evaluation of the 1 kJ modules during FY-82, and

e Development of the best candidate to a 30 kJ module with moderate pulse
rate capability by FY-85 or FY-86.

These developments may be compared with the near-term Nd glass and Co2 laser
projection given in Table XIII-A-3.

The next development tasks for lasers are associated with the efficient
and reliable generation of high power, rep-rated laser pulses. Master oscil-
lators which operate reliably with acceptable performance in the repetitive
pulsed mode need to be developed. The near term tasks associated with single
shot laser media handling must be solved for pulsed laser amplifiers. Here
the flow system and the medium reprocessing tasks become important. Thermal
control for these systems needs to be developed. Exciters which couple
efficiently with the Taser medium as well as efficient ways to extract the
high power pulse from the power amplifiers need to be developed.

b. Light Particle Beam

Electron accelerators provide one-step beam generation and acceleration
when a short high voltage electrical pulse is applied to the accelerator diode.
The temporal properties of the electron pulse leaving the diode are governed
by the detailed design of the diode/power supply combination. The attachment
of suitable ICF pulses (~10-30 manosecond pulse width) represents a signifi-
cant design problem for the overall system and will require R&D.

XI11-23
At present, electron diodes are being operated near the current levels
required for an ICF driver but the electron energies are too low. Diode-
power supply operation must also be extended to reach high power levels
(shorter pulses) and repetitively pulsed operation. Diode survivability
must also be developed for prolonged operation (109-10]05h0ts) and the ICF
nuclear environment.

Light ion accelerators use a diode generated electron beam to generate
the energetic light ion beam. Devices must be demonstrated with good con-
version efficiencies (electron to ion) if this is to become a viable driver
option. The ion generation techniques must also be developed to provide
adequate lifetime and maintainability.

To reach the pulse power and energy levels that will be required for
ICF, these accelerator systems are being assembled in parallel to reduce the
performance required of an individual accelerator module. This has been done
at Sandia on the recent electron machines and those planned for near term
construction (Table XIII-B-4).

TABLE XIII-B-4. Sandia Accelerators

Number of Peak Pulse Operation
Machine Modules Power Energy Date
Proto I 2 2 TW 12 KJ 1975
Proto II 8 TW 100 KJ 1977
EBFA I 36 30 TW 1T MJ 1982
EBFA 11 72 60 TW 2 MJ 1985

But the performance of the individual diodes must be improved beyond the
present State-of-the-Art if these drivers are to succeed in inertial con-
finement fusion.

The near term attainment of breakeven with these drivers will require
significant advances on single shot machines. This includes the improvement
of diode-power supply designs to yield shorter pulse lengths (10-20 ns).
Diode design must also accommodate the large power flow through the interface
between the power supply's pulse forming line (PFL)} and the evacuated beam

XIII-24
formation region. Improved materials and designs must be developed for this
interface region.

It is assumed for the purposes of this report that the light-ion diodes
are being developed to replace electron systems and that they will use the
exact same technology as the electron beam except the diode design. The
exact nature of the ion diode for ICF applications will depend on physics
experiments underway at this time at Sandia, NRL, University of Maryland,
I11inois, Cornell, Livermore/Berkeley and Ecole Polytechnique.

Major problems for diode development after breakeven are related to
repetitively pulsed operation and the nuclear environment. The diode must
withstand incident neutrons, ions and intense x-ray fields. Survivability
under commercial reactor conditions can only be simulated until the demon-
stration stage. Beam steering and transport techniques must be developed
to allow shielding and stand off of the diode from neutrons resulting from
the pellet implosions or the diode will have to withstand intense neutron
bombardment similar to the first wall in a Tokamak reactor. Diode replace-
ment by remote handling techniques, therefore, will have to be developed as
a part of the technology matrix.

Cooling methods to remove waste heat deposited in the diode by the beam
and by joule heating will have to be developed. For electron beams, thin
cathode foils or a fine mesh cathode screen will have to be cooled in high
current repetitively pulsed systems if electrostatic or magnetic cathode pro-
tection schemes are not successful in making the beam by-pass these elements.
If a portion of the beam is stopped in the support or mesh, this energy will
have to be removed. Furthermore, in e-beams the anode will have to be cooled
to remove the energy deposited from the plasma formed along the surface.
Likewise, the 1ight ion diode foil may have to be cooled.

c. Heavy lons

The development of heavy ion fusion drivers will not only benefit from the
existing machine experience but also from continuing accelerator development
programs. During the period covered by this plan, additional high energy
physics machines will be built in this country and abroad. Advances in machine
design, beam diagnostics, and control, etc. can be expected to occur.

XI11-25
In spite of the different operating regimes for fusion and high energy physics
machines it is likely that some results will be of significant value to the
heavy ion fusion program.

The fusion effort also stands to gain significant insights from recently
initiated programs to evaluate the feasibility of particle beam weapons. The
advances in this program could provide immediate benefits to the heavy ion
fusion program since their operating regimes are similar. Initial estimates
of weapons parameters include high peak currents (~10 kamperes), high peak
power (~10 TW), high pulse energy (~50-100kJ), and acceptable repetition rates
(5-50 pulses/sec.). The differences are that the military programs are exam-
ining the use of electrons and protons with energies of 0.5-1 GeV whereas the
fusion program needs range from 10-100 GeV. Major technical areas included
in the present program that would relate to the heavy ion fusion are beam
propagation physics, accelerator technology, and power supply subsystems
(switches, energy storage, energy generation). A major barrier for incorpor-
ating these gains into the commercial heavy ion fusion program could result
from classification of particle beam program advances.

High energy physics accelerators perform sequential manipulations of the
beam over large system lengths (several kilometers for HIG systems). A1l
systems proposed for HIF include 1) an ion source and preacceleration stage,
2) a voltage gain accelerator, and 3) a device or system to increase the
instantaneous current levels (pulse compression). The voltage and current
gain operations may be provided by one or more individual accelerator sub-
systems. The sequential nature of beam manipulation with these systems makes
them suitable for performance upgrades by adding stages to an existing machine.
This has the potential to reduce costs by requiring only one phased machine
instead of several.

d. ICF Fusion Targets

In practice bare D-T pellets or pellets with single container materials
will not be suitable to obtain the required compression and heating. Future
pellets will probably consist of a number of concentric spherical regions of

3)

outside surface of these layers and its interfaces between layers will be

different materials and will have diameters in the millimeter range.‘ The

XIII-26
smooth enough to prevent Rayleigh-Taylor stabilities (defect height less than
100A and spatial wavelength of 50 microns or less). The development of eco-
nomical fabrication techniques of such targets will require a considerable
efforts.

5. ICF R&D Facilities

In addition to the development of lasers as drivers for Inertial Confine-
ment Fusion (ICF), the current Office of Inertial Fusion (OIF) program plan
includes light-particle beam (electron beam) (EBFAII) and heavy jon beam
scientific feasibility demonstrations by ]985 whose purposes are described in
Table XIII-B-5 and having a schedule as indicated in Figure XIII-B-4. The
next generation of engineering development facilities and reactors, as indi-
cated in this figure, would not allow the scheduling of operation of an
ENgineering Test Facility equivalent to the magnetic fusion ETF until 1993,
This is principally due to the fact that the ICF program presently does not
plan to conduct any significant engineering development activities until after
scientific feasibility is demonstrated in each of their driver options by 1985,
Thus, as seen in Figure XIII-B-4, the ICF Engineering Test Facility (ETF) is
scheduled to operate (c. 1993), approximately one year after the magnetic
fusion scheduled ETF (See Figure XIII-B-1)} which is part of a program that
has already proceeded with significant engineering development (See Figure
XIII-B-2).

The proposed schedule for the ICF Fusion Pilot Plan and Prototype Fusion
Power Plant which would precede the first commercial demonstration could
readily accommodate the equivalent hybrid facilities on the same schedule,
viz., the EHR and PHR. It should be noted that this ICF schedule of facilities
allows 2-3 years of operational experience and data to impact the final con-
struction design while the magnetic fusion schedule makes for 1ittle if any .
of such allowance.

XI1I-27
TABLE XIITI-B-5.

Facility
ANTARES
NOVA 11

EBFA I1I
Advanced Laser
Heavy Ion Beam

Systems Integration Facility

Single-Pulse Target Facility
Engineering Test Facility

Materials Test Facility

Fusion Pilot Plant

Prototype Fusion Power Plant

Objectives of Major ICF Facilities

Year of
Operation

1982
1984

1985

1985

1985

1988

1989

1994

1997

2000

XII1-28

Objective

Achieve pellet gains of 1

Achieve pellet gains of
20 or more

Achieve pellet gains of
up to 10

Repetitive operation at
30 kd

Demonstrate heavy ion
beam driver feasibility

Driver pellet targeting

- Pulse power supply testing

Driver-module testing
Beam propogation studies

Commercial pellet development
Reactor component testing

Reactor systems qualification
in pulsed nuclear environment

Materials qualification
Pulsed radiation effects
testing

Confirm prototype plant
technology
Electric power production

Safe reliable operation

Scaleable to commercial size
ANTARES

6Z-111X

CURRENT NOVA {1 [ DESIGN

FKE‘}N"TEIES EBFA 1 CONSTRUCTION, MODIFICATION
ADVANCED LASER B OPERATION
HEAVY ION BEAM C—1 NO SPECIFIED TIME LIMIT
SYSTEMS INTEGRATION FACILITY

NEXT SINGLE-PULSE TARGET FACILITY
GENERAT ION ENGINEERING TEST FACILITY
FACILITIES MATERIALS TEST FACILITY

FUSION PILOT PLANT
PROTOTYPE FUSION POWER PLANT

»

‘L

2000 05 10

DATE

FIGURE XIII-B-4. Inertial Confinement Fusion Facilities Schedule

C. PRESENT STATUS OF BLANKET ENGINEERING

1. Neutronics Design

Hybrid concepts have thus far been subjected mainly to survey type neu-
tronic analysis with specific requirements in the calculations, nuclear data,
and experimental areas. Detailed calculations have not been made, in part
because of the uncertainties associated with structural material choice and
requirements which require definition from hydraulic and engineering analyses
and the exact fluence limitation of a given design. Costs of conducting
detailed analyses is another factor. Existing calculations of hybrid concepts,
with few exceptions, have not been performed with the full sophistication
available to fission reactor designers and with little attention to nuclear
data problems. The experimental basis for the correlation of calculations
is sparse and no experimental neutronics program on hybrid concepts presently
exists.

a. Nuclear Data

The only nuclear data uncertainties for hybrid concepts which have been

studied have been for the 238

U nucleus. No nuclear data problems have been
considered for 232Th or for the fissile isotopes and the minor isotopes which
impact on fuel cycle problems. Complete evaluated nuclear data files have
been prepared for essentially all of the actinide nuclei for the next version
(V) of the U.S. DOE Evaluated Nuclear Data File (ENDF/B). These files have

- not yet been released for use and no assessment has been made of their uncer-

tainties and adequacy for hybrid neutronics analysis.

b. Calculational Analysis

Most existing hybrid neutronic calculations are very unsophisticated.
Most have not included resonance self shielding or temperature effects. Only

a few calculations exist with neutron thermalization and associated temperature

232,, 236 238

dependence. With one exception the minor isotopes U, Pu and Pu and

their impacts on materials handling, operations, biological radiation dose,
and safeqguards have not been evaluated. Calculations of the fast-thermal

233U refresh mode hybrid have been extremely cursory at this time. The

XITI-30
question of fission criticality safety has been subjected to little rigorous
analysis. There are very few studies to determine optimum fuel management
steps for improved blanket performance. Inertial confinement fusion systems
have not been evaluated for performance with high density pellets with their
associated neutron spectrum and intensity characteristics. The effect on
tritium breeding requirements needs to be examined.

c. Experiments

No integral measurements experimental program exists for the hybrid
concept. Some attempts have been made to define areas of microscopic
neutron cross-sections for hybrid applications where improved experimental
data are needed. There is no established experimental program supported to
supply these needs. A single integral experiment performed some 20 years
ago with a 14 MeV neutron source in a natural uranium assembly has constituted
the basic criterion for hybrid neutronics performance. A1l calculational
comparisons except one, however, have to date incorrectly compared calculated
with experimentally measured quantities.

2. Thermal and Mechanical Design

The thermal and mechanical design of the blanket is more dependent on
the fusion driver than on the type of fission blanket selected. Configur-
ations are laid out to fit the geometry of the fusion driver and to facili-
tate the removal of blanket material. This will be required of all designs
depending on the fission fuel cycle being considered and exposure capability
of the fission fuel form and structural components utilized in the design.
It is important in conceptual design studies to establish the required blanket
configuration early in the study. Preliminary thermal-hydraulic and struc-
tural analyses must be carried out in conjunction with neutronic survey
calculations. The neutronic performance is sensitive to the amount and type
of structural material required in the blanket as noted previously, particu-
larly for thermal and fast-thermal lattices. Knowing the blanket configur-
ation is also important for estimating the integrated hybrid performance
since the overall performance is sensitive to the fraction of total fusion
neutrons utilized by the blanket.

XIII-31
a. Fuel Form

The selection of a fission fuel form for any given blanket has been
based primarily on neutronic performance, but the candidate forms have
been based on developed fission reactor technology. Fusion-fission hybrid
designers utilizing near-term fusion drivers, i.e., those with high recir-
culating power and low plasma, have selected advanced fission fuel forms
and cladding for their designs. Those utilizing fusion drivers having a
high gain (e.g., ignition tokamaks), however, can obtain performance para-
meters of commercial interest with blankets utilizing near-term fission
fuel forms and cladding such as UC clad in stainless steel.

b. Blanket Coolants

Selection of hybrid blanket coolants has been based on several factors:

Compatibility with neutronic requirements of the blanket

Status of power conversion system components

Availability of design analysis methods and supportive data bases
Compatibility with fuel form, cladding and structural materials
Compatibility with tritium processing requirements

Knowledge of magnetic field effects

Ability to predict safety performance

The selected coolant must be compatible with ultimate transfer of heat to a
modern steam system to maintain reasonable power conversion efficiency within
the temperature limitations of available blanket structural materials. In

all the areas of technology previously mentioned, we know the most about water
as a coolant. Extensive R&D in the LWR program has developed an adequate base
and design methods to predict water-cooled blanket performance. However,
water has not been considered as a blanket coolant to date because it is very
difficult if not impossible to remove tritium from water. In LWRs tritium
releases outside the plant are controlled simply by 1imiting the generation

of tritium. Impurities (Li) in the core are reduced to levels which Timit

the tritium production to amounts than can be released from the plant.

The HTGR and German Gas Cooled Reactor programs have developed and
demonstrated helium cooled power conversion system technology. Helijum is

XI1I-32
compatible with all structural materials with the exception of refractory
metals and alloys. The impurity levels attainable in real systems result
in corrosion problems for the refractories.

To get adequate heat transfer and transport properties, helium systems
have to be operated at relatively high pressures (50 to 70 atms.). 1In the
complex geometries of hybrid blankets this results in a requirement for lots
of structural material fractions which reduces neutronic performance (i.e.,
parasitic absorption of the neutrons)., Where cladding and structural mate-
rials are stainless steel, helium-cooled systems yield 30% power conversion
efficiency. If higher temperature alloys (TZM, Inconel, etc.) are used,
efficiencies approaching 40% are possible. Helium has good neutronic
properties with no anticipated MHD or corrosion enhancement effects in
magnetic fields. Hence, it has been a nearly unanimous choice of designers
for use in Tokamak and Mirror hybrid blankets.

The LMFBR program is developing data and system components for Na cooled
systems. The major uncertainties in Na cooled systems are the MHD effects
in rapidly changing high magnetic fields and the effects of magnetic fields
on corrosion and mass transport rates. Due to enhanced heat transfer, higher
sodium temperatures can be achieved with stainless steel structural materials
and thus power conversion efficiencies near 40% can be acheived without the
use of high temperature alloys. The LMFBR program is also developing an
extensive safety related data base for Na coolant. These data will be directly
applicable to assessing hybrid safety problems.

Few hybrid designers to date have not proposed using Li as a coolant,
Although it is attractive neutronically for producing tritium, the technology
base for Li leaves uncertainties. Li appears to be more corrosive than Na
and hence operating temperatures must be lower (50°C) to be compatible with
stainless steel, resulting in lTower power conversion efficiency. The increased
“corrosion and mass transport rates result in uncertainty in the applicability
of current Na power conversion system components.

Because 1iquid metals can be used at low pressures, they result in low
structural material requirement. Where magnetic field effects are not impor-
tant (laser applications) designers have proposed using both Na and Li as
coolants, thus maximizing the use of R&D benefits from the LMFBR program.

~ XITI-33
If candidate coolants are ranked by the available technology base, they
would fall in the following order:

e Water coolant

e Helium coolant
e Sodium coolant
® |ithium coolant

c. Design Analysis

Methods currently being developed for the LWR safety program and LMFBR
program should be adequate for analyzing the response of proposed fuel forms
to start up transients or pulsed power operation. Ultimately experimental
verification will be needed.

The calculational method for heat transfer and fluid flow, developed by
the fission reactor programs, are adequate for conceptual hybrid reactor
blanket designs. However, detailed design and safety analyses of start-up,
pulsed operation and abnormal transients are going to require much closer
coupling of thermal and mechanical analysis methods than now exists for both
fuel and structures.

The structural analysis methods currently employed by designers (BOSOR 4,
AXISOL, etc.) will require careful modeling by experienced structural analysts
to adequately predict the stresses imposed on the complex modular and coolant
header structures resulting from the Tokamak and Mirror hybrid conceptual
design studies. None of the design teams have been adequately funded to date
to take a design and set up all the structural calculations. This should be
done to test the applicability of the current analytical methods as well as
to give a preliminary assessment of the important design problems.

XITI-34
D. BLANKET RD&D REQUIREMENTS

1. Fission

The specific fission research and development needs for fusion-fission
hybrid blanket designs will depend to a large extent on the fuel form and
cladding, coolant, structural complexity, and operational characteristics
of the concept(s) to be developed. Conceptual design studies to date have
not been performed to the depth to have identified a reasonable list of
the major R&D requirements that will be necessary for development of any
given concept. It is therefore most appropriate to identify the very
general requirements that are resulting from preliminary choices being
made by conceptual designers.

Initial hybrid studies concentrated on fuels which have a good techno-
logical base from the fission reactor program. However, the more advanced
studies have moved to advanced fuels and cladding that add uncertainty to
the fabrication, irradiation behavior and reprocessing technologies involved.
It is becoming increasingly apparent that hybrids with performance of commercial
interest will require considerable development of the fission fuel form and
associated technologies if near-term fusion drivers are utilized. Some
alternative fuels considered are part of the current U.S. fission reactor
research and development program while others are not. Nevertheless, the
conditions imposed by 14 MeV neutrons and pulsed operation will require con-
siderable performance verification since current transient experiments on
oxide fuel show very limited ability to withstand cyclic operation. Economic-
ally attractive blanket designs for all hybrid concepts tend to select fuel
forms which maximize the heavy metal density in the blanket lattice. This
means a carbide or metallic fuel form. Assuming the LMFBR Program will
develop the basic irradiation performance data for these fuel concepts, the
question of how do 14 MeV neutrons, pulsed operation, alternate high tempera-
ture cladding, and severe power gradients affect the applicability of the
basic data will have to be answered.

Methods for predicting fuel and clad transient response will have to
be adapted from the LWR and LMFBR safety programs to the problems of
predicting hybrid fuel response to rapid transients and pulsed operation.
Experimental verification of these methods will have to be performed.

XII1-35
2. Neutronics

The neutronics RD&D requirements fall in four general areas: nuclear data,
analytical methods, conceptual design studies, and integral experiments,

The nuclear data area includes the development of evaluated nuclear data
files, experimental measurements of microscopic nuclear cross-sectional data,
and the function of nuclear data centers, The development, assessment, and
improvement of the evaluated nuclear data files require a continuing effort
throughout the hybrid development, preferably carried out in conjunction with
an experimental measurements program and conceptual design studies, The
experimental cross-section data program requires the partial support of one
or more measurements facilities with continuous source energy capability to
14 MeV neutrons.

The RD&D effort in analytical methods consists of the development and
maintenance of verified computer codes to handle radiation deep penetration
problems with complicated geometries, voids, and anisotropies. It also
requires the melding of fission reactor core physics and burnup codes along
with the corresponding required nuclear data libraries.

The RD&D requirements for conceptual design studies require the application
of these codes and data in evaluations of conceptual hybrid designs.. The studies
required should emphasize performance optimization and full attention to detail
of all aspects of the nuclear fuel cycle.

The integral experiments require an improved high-density 14 MeV neutron
source facility with associated nuclear measurements capability. Integral
measurements for verification of design studies and correlation with theoretical
methods and nuclear data are required. The measurements program should proceed
for simple homogeneous experiments through the complicated heterogenous blanket
systems required for most hybrid concepts.

XIII-36
A1l of the above development could be implemented in the proposed blanket

and shield experimental facility of Figure XIII-B-2 which presumably has a
source of 14 MeV neutrons and capabilities for simulating the conditions of
hybrid blanket neutronic exposures as well as thermal, mechanical and fuel
testing. This facility could therefore impact the EHR and PHR, However, in
preparation for such a facility and in order to provide design data for the
HEF, hybrid blanket and fuel development would have to be initiated earlier

in more modest laboratory facilities to impact those designs.

XII1-37
E. POSSIBLE HYBRID RD&D PROGRAM

1. Program

The formulation of a definitive Research, Development and Demonstration
Program of the fusion-fission hybrid energy system concept would be more
formidable a task than could be accomplished during the course of this study.
However, based upon the current status and RD&D requirements as reviewed in
the preceding sections for magnetic and inertial fusion drivers, as well as
hybrid blanket engineering and fission fuel, a Hybrid RD&D program can be
proposed which could serve as a framework for future RD&D assessments and
planning.

The advance technological requirements for commercializing hybrid
reactors, some of which have been addressed in the preceding sections, are
tabulated in Table XIII-E-1. 1In order to implement this development, a series
of integrated RD&D facilities and projects may be planned beyond alternate
magnetic and inertial fusion driver development, selection and demonstration
of scientific feasibility as indicated by the schedules of Fiqures XIII-B-1
and XIIT-B-4. Assuming such driver development selection and demonstration
will proceed through 1985, a proposed parallel schedule of hybrid development
facilities with magnetic and inertial fusion drivers and hybrid blankets is
shown in Figure XIII-E-1.

In this program parallel Hybrid Experimental Facilities (HEF) and Experi-
mental Hybrid Reactors (EHR) would be constructed with both magnetic and
inertial fusion drivers which have survived their respective fusion development,
selection and demonstratior program. After operational experience of both
magnetic and inertial EHRs, a decision will be made as to which fusion driver
will be selected for the Prototype Hybrid Reactor (PHR) of near or full com-
mercial size. Preceding and in parallel with the dual HEF and EHR facilities,

a program of facilities to conduct hybrid blanket and fuel development will
be required.

XITI-38
TABLE XIII-E-1. Hybrid Reactor Technological
Advance Requirements

Plant Components

Modest improvement in nerformance

or size from present knowledge

Modest improvement in performance or

size and modified confiquration/

modified confiquration/application
application

No new knowledge required
Contemporary technology with

Major improvement in performance or
size and modified configuration/

Major improvement in performance or
application

size from present knowledge

required to meet system require-

New technology {e.g., materials)
ments

New technology and modified design

required

Entirely new concept requiring new

technoloqy and new design

N

< £ e+ 0 35 O T O 3

Nuclear Fuel

Driver control systems
Driver vessel

Blanket support structure

Blanket vessel internals including
shielding, ducting, control rod
guides, baffles, etc.

Primary coolant pumps and auxiliary
systems

Primary coolant chemistry/
radiochemistry control

Primary system heat exchangers
Reactor instrumentation

Emergency core cooling/safe
shutdown systems

Containment, containment cleanup
systems and effluent control systems

Other accident mitigating systems,
j.e., plant protection systems

On-site fuel handling storaae/
shipping equipment

Main turbine

Other critical components, if any
Balance of plant components

Fuel raw material

Fresh/recycle fuel fabrication
Reprocessing

O0ff-site fuel storage/disposal
Radioactive waste disposal

Special nuclear safequards and
protection

New supporting technologies

Fission
Fusion
Purex
Civex
Thorex

.

DB W -

XIII-39

| AU

Op-I1IX

&
>
el
w

79180 |81 82| 83] 84] 85| 86] 87] 88] 89 [90] 919293 [0a[ 95| 96] 97 [98] 99| 00 |01 | 02| 03] 04 05 | 0607 ] 08 [ o9 [ 10]

MAGNETIC
FUSION DRIVER

oo MM

4

EA NN

BLANKET MOD ULE Iy

DEVELOPMENT

-

SN

m N,
z HYBRID Y *\’% PROTOTYPE
a FACILITY RN f 5 REACTOR
g | } \
S Iz
HYBRID FUEL
DEVELOPMENT HYBRID ' [ CONCEPTUAL DESIGN
AND TESTING EXPERIMENTAL EXPERIMENTAL \\ 1 CONSTRUCTION
& FACIUTY\ Y SET«%:RTI(?R \\ OPERATI ON
=
= SPTF
SIF
veaRs| 79] 80] 81 82 23] 24| 85] 86| 87] s8] 89] 90] 91] 92| 93] 94] 95| 96} 97| | 99 00 ] 01] 02] 03[ 0a | 5 | 06 07| 68| 0] 10]

FIGURE XITI-E-1.

Hybrid Development Facilities Scheduie

a. Prototype Hybrid Reactor

The PHR will be a near or full commercial sized hybrid system with all
integrated components prototypical of those to be used in commercial systems.
Its driver selectiorn (magnetic or inertial) may determine the plant size. It
would demonstrate electric power and fissile fuel production in a reliable,
efficient, maintainable, integrated system which is licensed and operating on
a utility grid. This will require high plant efficiency and availability of
a plant in the 500-1000 MWe range producing 1000-2000 kg/yr of fissile fuel.
The construction and operation costs should be able to be readily extrapolated
to commercial hybrid plants.

b. Hybrid Blanket

The hybrid blanket facilities include parallel facilities to conduct
blanket module coolant and fuel development and testing in thermally and
mechanically simulated and fission reactor experimental enviroments. Such
development will support the Hybrid Blanket Facility (HBF) which will have a
dedicated 14 MeV neutron source of sufficient strength, fluence and target
volume to perform single modular hybrid blanket experiments and testing.

The HBF will be a long-term facility which together with those support
facilities will qualify blanket module and fuel designs for testing in the
HRE, EHR and eventually the PHR facilities. '

XI11-41
2. Facilities

a. Magnetic Fusion

The HEF would be the hybrid equivalent of the magnetic fusion ETF. This
would require an appropriate minimum driver size to produce reactor grade
plasmas and having a sufficient duty cycle to performing engineering tests
of the various driver subsystem components including first wall, blanket
and shield, superconducting magnets, heating and fueling systems. It would
not have to breed tritium; however, it must be capable of performing tritium
breeding experiments in appropriate blanket modules. Its hybrid engineering
capabilities must facilitate the in-situ experiments of various hybrid blanket
module, fuel and coolant selections as initially developed in the hybrid
blanket facilities.

The EHR would be the first power and fuel producing demonstration of the
hybrid system. It will have many scaled down driver components prototypical
of a commercialized sized hybrid. It will be capable of producing significant
power (100 to 300 MWe) and fissile fuel (100 to 1000 kg/yr) while simultaneously
breeding tritium in a self consistent fusion fuel system. This will require a
fully integrated reactor system having a reasonable duty cycle and plant factor.

b. Inertial

The Single Pulse Target Facility (SPTF) will be used for commercial pellet
and 1imited reactor component development. It would have a powerful driver
capable of producing 0.1-1.0 MJ per pulse operating in the single, discrete
pulse mode for pellets having gains 10-100. The System Integration Facility
(SIF) would develop and integrate high-repetition-rate subsystems for commercial
reactor operation with driving pellets. It could conceivably be the initial
phase of the inertial HRE with a 1 Hz, 0.1-0.2 MJ driver/targeting system with
reasonable duty cycle to perform hybrid blanket modular experiments. Pellet
manufacture and tritium breeding would not be required of this system although
it should have the capability of performing tritium breeding modular experiments
and it may require its own pellet factory.

The inertial EHR would be similar in objectives to its magnetic counterpart;
however, it may be significantly smaller in power and fuel producing capability
due to the modularity of inertial fusion systems.

XII1-42
3. Funding Reauirements

Estimates of the total expenditures for develooment, design and
construction of the hybrid program facilities have been made and are given in
Table XIII-E-2. These estimates are based upon the normalized cost estimating
procedures used in Section IX, which have been developed by PNL for OFE and
OLF, as well as upon the cost estimates developed by PNL for the ICF facilities
in the Engineering Development Program Plan.

One might note from these costs and the schedule of Figure XIII-E-1 that

the ICF, HRE and EHR facilities require somewhat less funds for design develop-
ment and construction, as well as short construction periods, than their magnetic

fusion counterparts. This is principally due to the fact that ICF hybrid
related and some common fusion system components will piagyback the magnetic
facility development. In addition, the equivalent ICF hybrid system will
generally be of smaller size than the magnetic system because of the modularity
of ICF systems and their potential rep rate and target gain flexibilities. It
should be noted that the cost estimates do not include the operational and
testing cost associated with the development program which may require an
additional $3 to $5 billion to commercialization. Also, the Federal Government
funded PHR facility is significantly less expensive than the commercial hybrid
systems costed in Section IX since they are unoptimized advanced developed
full-scale commercial systems paid for with private capital. It is expected
that an optimization of the performance and cost of these systems would
implement cost reduction opportunities to achieve the same performance at a

15 to 20% cost reduction.

XITI-43
TABLE XIII-E-2. Hybrid Facility Cost Estimates

XITI-44

Development,
Desian and
Operational Construction
Facility Date (FY) Costs (M 1978)

Magnetic Fusion

HER 1989 800

EHR 1998 1200
Inertial Fusion

SIF 1988 100

SPTF 1989 500

HFF 1993 600

EHR 2000 1000
Blanket

Blanket Module and

Coolant Development 1984 200

Hybrid Fuel

Development and Testing 1983 200

Hybrid Blanket Facility 1988 400
Prototype Hybrid Reactor 2010 2000

Total $7000M
SECTION XTIT REFERENCES

Magnetic Fusion Programs Summary Document - FY-1980 HCP/73168-01, TRW,
Inc., Redundo Beach, CA, April 1979.

R. E. Aamodt, et al., Assessment of the Tokamak Confinement Data Base.
EPRI ER-714 Electric Power Research Institute, Palo Alto, CA, March 1978.

J. H. Nuckolls, "Inertial Confinement Fusion Targets," Inertial Confine-
ment Fusion. Optical Society of American, page TuA-5-1, Washington, DC,
1978. p. TuA-5-1.

XII1-45
APPENDIX A
CAPITAL INVESTMENT COST ESTIMATES
Laser Inertial Confinement Hybrid Reactor Capital Costs ($106)a

Pu Recycle/
Once Through

Account Number

20
21

22

23

24

25

27

91

92

93

21.
21.
21.
21.
21.
21.

22.

22.
22.
2z,
22.
22,
22.
22.
22.

22,
22,

01
01
01
01
01
01
0
01

02.
02.

.01
.02
.03
.04
.05
.06
.08
.09

01
02

(a)

June 1978 dollars

LAND AND LAND RIGHTS

STRUETURES AND SITE FACILITIES

Site Improvements and Facilities
Reacter Building

Turbine Building

Miscellaneous Buildings

Spare Parts Allowance (0.5%)
Contingency Allowance {20%)

REACTOR PLANT EQUIPMENT

Reactor Equipment
Blanket and First Wall
Shield
Magnets
Supplemental Heating
Primary Support and Structure
Reactor Vacuum Systems
Impurity Control
Direct Energy Converter
Main Heat Transfer and Transport Systems
Primary Coolant System
Intermediate Coolant System
Auxiliary Cooling Systems
Radiocactive Waste Treatment and Disposal
Fuel Handling and Storage Systems
Other Reactor Plant Equipment
Instrumentation and Control
Spart Parts Allowance {1%)
Contingency Allowance (30%)

TURBINE PLANT EQUIPMENT

Turbine-Generators

Main Steam (or other Fluid) System
Heat Rejection System

Cther Turbine Plant Equipment
Spare Part Allowance {1%)
Contingency Allowance {20%)

ELECTRIC PLANT EQUIPMENT

Spare Part Allowance (0.5%)
Contingency Allowance {(20%)

MISCELLANEQUS PLANT EQUIPMENT

Transportation and Lifting Equipment
Air and Water Service

Other Plant Equipment

Spare Parts Allowance (1%)
Contingency Allowance (20%)

LASER SYSTEM EQUIPMENT
Spare Parts Allowance {17)

Contingency Allowance (30°)
Total Direct Cost

CONSTRUCTION FACILITIES, EQUIP-
MENT AND SERVICES (157)

ENGIMEERING AND COMNSTRUCTION
MANAGEMENT SERVICES (157)

OTHER COSTS (57)
Total Indirect Cost

TOTAL CAPITAL COST

A-1

2.5

64.09

195.50

1.95
58.65
256.10

¢26.38

226.38
75.46

1509.23

525.23
2037.46
Laser Inertial Confinement Hybrid Reactor Capital

Account Number

20
21

22

23

24

25

27

9

92

93

21,

21
21
21
21
21

22,

0l

.02
.03
.06
.98
.99

01

22
22

22
22

22,
22.

.01
.01
22,
22.
22,
22.
.0
.01,

o
01
01
01

02.
02.

.01
.02
.03
.04
.05
.06
.08

09

01
02

(a)

June 1978 dollars

LAND AND LAND RIGHTS

STRUCTURES AND SITE FACILITIES

Site Improvements and Facilities
Reactor Building

Turbine Building

Miscellaneous Buildings

Spare Parts Allowance (0.5%)
Contingency Allowance (20%)

REACTOR PLANT EQUIPMENT

Reactor Equipment
Blanket and First Wall
Shield
Magnets
Supplemental Heating
Primary Support and Structure
Reactor Vacuum Systems
Impurity Control
Direct Energy Converter
Main Heat Transfer and Transport Systems
Primary Coolant System
Intermediate Coolant System
Auxiliary Cooling Systems
Radioactive Waste Treatment and Disposal
Fuel Handling and Storage Systems
Other Reactor Plant Equipment
Instrumentation and Control
Spart Parts Allowance (1%)
Contingency Allowance (30%)

TURBINE PLANT EQUIPMENT

Turbine-Generators

Main Steam (or other Fluid) System
Heat Rejection System

Other Turbine Plant Equipment
Spare Part Allowance (1%)
Contingency Allowance (20%)

ELECTRIC PLANT EQUIPMENT

Spare Part Allowance (0.5%)
Contingency Allowance (20%)

MISCELLANEQUS PLANT EQUIPMENT

Transportation and Lifting Equipment
Air and Water Service

Other Plant Equipment

Spare Parts Allowance (1%)
Contingency Allowance (20%)

LASER SYSTEM EQUIPMENT

Spare Parts Allowance (1%)
Contingency Allowance (30%)

Total Direct Cost

CONSTRUCTION FACILITIES, EQUIP-
MENT AND SERVICES (15%)

ENGINEERING AND COMSTRUCTION
MANAGEMENT SERVICES {15%)

OTHER COSTS (5%)
Total Indirect Cost

TOTAL CAPITAL COST

A-2

Costs ($106)a

Th-Pu
Catalyst
2.5

23.65
n2.42
72.41
95.87
1.13
45.37
273,135

96.71

195.50

1.95
58.65
256.10

308.42

308.42
102.81

2056.15

719.65
2775.80
Laser Inertial Confinement Hybrid Reactor

Account Number

20
21

22

23

24

25

27

91

92

73

22

.G
.G2
.G3
.G6
.98
.99

.01

22,
22.
22.

22

22.
22.
22.

22

22.

-
(4

e
e
01.
.01.
01,
01.
01.
.01.

0z.
.02,

LAND AND LAND RIGHTS

STRUCTURES AND SITE FACILITIES

Site Improvements and Facilities
Reactor Building

Turbine Building

Miscellaneous Buildings

Spare Parts Allowance (0.5%)
Contingency Allowance (20%)

REACTOR PLANT EQUIPMENT

Reactor Equipment
Blanket and First Wall
Shield
Magnets
Supplemental Heating
Primary Support and Structure
Reactor Vacuum Systems
Impurity Control
Direct Energy Converter

Capital Costs ($106)a

Refresh
Cycle
2.5

14,32
49.90
13.57
58.04
0.69
27.67
166.68

58.55

Main Heat Transfer and Transport Systems

Primary Coolant System
Intermediate Coolant System
Auxitiary Cooling Systems

Radioactive Waste Treatment and Disposal

Fuel Handling and Storage Systems

Other Reactor Plant Equipment
Instrumentation and Control
Spart Parts Allowance {1%)
Contingency Allowance {30%)

TURBINE PLANT EQUIPMENT

Turbine-Generators

Main Steam {or other Fluid) System

Heat Rejection System

Other Turbine Plant Equipment
Spare Part Altowance {1%)
Contingency Allowance (20%)

ELECTRIC PLANT EQUIPMENT

Spare Part Allowance {0.5%)
Contingency Allowance {20%)

MISCELLANEOUS PLANT EQUIPMENT

(a)

June 1978 dollars

Transportation and Lifting Equipment

Air and Water Service
Other Plant Equipment
Spare Parts Allowance {1%)
Contingency Allowance {20%)

LASER SYSTEM EQUIPMENT

Spare Parts Allowance {1%)
Contingency Allowance {30%)

Total Direct Cost

CONSTRUCTION FACILITIES, EQUIP-
MENT AND SERVICES (157)

ENGINEERING AND CONSTRUCTION
MANAGEMENT SERVICES (15%)

OTHER €OSTS (5%)
Total Indirect Cost

TOTAL CAPITAL COST

A-3

195.50

1.95
58.65
256.10

1216.01
182.40

182.40
60.80
425.60
1641.62
Classical Mirror Hybrid Reactor Capital

Account Number

20
21

22

23

24

25

91

92

93

22.

24,
.99

24

25.
25,
25.
25.
25.

.01
.02
.03
.06
.98
.99

01

98

01
02
07
98
99

22.
22.
22.
22,
22.
22.
22.
22.

22.
22.

.01
.02
.03
.04
.05
.06
.08
.09

.01
.02

(a) June 1978 dollars

LAND AND LAND RIGHTS

STRUCTURES AND SITE FACILITIES

Site Improvements and Facilities
Reactor Building

Turbine Building

Miscellaneous Buildings

Spare Parts Allowance {0.5%)
Contingency Allowance {20%)

REACTCR PLANT EQUIPMENT

Reactor Equipment
Blanket and First Wall
Shield
Magnets
Supplemental Heating
Primary Support and Structure
Reactor Vacuum Systems
Impurity Control
Direct Energy Converter

Costs ($106)a

Pu Recycle/
Once Through
2.5

3.66
23.46
11.60
18.17

0.30
11.88

_I1.57

64.55
19.89
119.38
142.22
263.38
253.81

67.30

Main Heat Transfer and Transport Systems

Primary Coolant System
Intermediate Coolant System
Auxiliary Cooling Systems

77.34
64.45
53.73

Radioactive Waste Treatment and Disposal

Fuel Handling and Storage Systems
Other Reactor Plant Equipment
Instrumentation and Control

Spart Parts Allowance (1%)
Contingency Allowance {30%)

TURBIRE PLANT EQUIPMENT

Turbine-Generators

Main Steam (or other Fluid) System
Heat Rejection System

Other Turbine Plant Equipment
Spare Part Allowance {1%)
Contingency Allowance (20%)

ELECTRIC PLANT EQUIPMENT

Spare Part Allowance (C.5%)
Contingency Allowance (20%)

MISCELLANEOUS PLANT EQUIPMENT

Transportation and Lifting Equipment
Air and Water Service

Other Plant Equipment

Spare Parts Allowance (1%)
Contingency Allowance (20%)

Total Direct Cost

CONSTRUCTION FACILITIES, EQUIP-
MENT AND SERVICES {15%)

ENGINEERING AND CONSTRUCTION
MANAGEMENT SERVICES {15%)

OTHER COSTS (5%)
Total Indirect Cost

TOTAL CAPITAL COST

46.38

8.28
11.80
354.21
1546.72

14.04
93.74
26.94
56.59
1.9
38.26
231.48

24.52

0.12
4.90
29.54

1.42
4,23
14.62
c.20
4.05
24.52

285.57

285.57

95.19

1903.83

666.33
2570.16
Account Number

Classical Mirror Hybrid Reactor Capital Costs ($106)a

20
21

22

23

24

25

91

92

93

2z,

01

22,
22.
22.
22.
22.
22.
22.
22.

22.
22.

{a) Jdune 1978 dollars

01

01
01
01
01
01
01

02

.01
61.

02

.03
.04
.05
.06
.08
.09

.01
02.

02

LAND AND LAND RIGHTS

STRUCTURES AND SITE FACILITIES

Site Improvements and Facilities
Reactor Building

Turbine Building

Miscellaneous Buiidings

Spare Parts Allowance (0.5%)
Contingency Allowance (20%)

REACTOR PLANT EQUIPMENT

Reactor Equipment
Blanket and First Wall
Shield
Magnets
Supplemental Heating
Primary Support and Structure
Reactor Vacuum Systems
Impurity Control
Direct Energy Converter

Main Heat Transfer and Transport Systems

Primary Coolant System
Intermediate Coolant System
Auxiliary Cooling Systems

Radioactive Waste Treatment and Disposal

Fuel Handling and Storage Systems
Other Reactor Plant Equipment
Instrumentation and Control

Spart Parts Allowance (1%)
Contingency Allowance {30%)

TURBINE PLANT EQUIPMENT

Turbine-Generators

Main Steam (or other Fluid) System
Heat Rejection System

Other Turbine Plant Equipment
Spare Part Alilowance (1%)
Contingency Allowance (20%)

ELECTRIC PLANT EQUIPMENT

Spare Part Allowance (0.5%)
Contingency Allowance (20%)

MISCELLANEQUS PLANT EQUIPMENT

Transportation and Lifting Equipment
Air and Water Service

Other Plant Equipment

Spare Parts Allowance (1%)
Contingency Allowance (20%)

Total Direct Cost

CONSTRUCTION FACILITIES, EQUIP-
MENT AND SERVICES (15%)

ENGINEERING AND CONSTRUCTION
MANAGEMENT SERVICES (15%)

OTHER COSTS (5%)
Jotal Indirect Cost

TOTAL CAPITAL COST

A-5

Th-Pu
Catalyst

2.5

332.49

710.80

2215.98

775.60

2591.58
Classical Mirror Hybrid Reactor Capital Costs ($106)a

Account Number

20
21

22

23

24

25

91

92

93

21.
21.
21.
21.
21.
21.

22

.01

.01.
.01.
.01.
.01.
.01.
.01.
.01.
.01.

.02.
.02.

(a) June 1978 dollars

LAND AND LAND RIGHTS

STRUCTURES AND SITE FACILITIES

Site Improvements and fFacilities
Reactor Building

Turbine Building

Miscellaneous Buildings

Spare Parts Allowance (0.5%)
Contingency Allowance (20%)

REACTOR PLANT EQUIPMENT

Reactor Equipment
Blanket and First Wall
Shield
Magnets
Supplemental Heating
Primary Support and Structure
Reactor Vacuum Systems
Impurity Control
Direct Energy Converter
Main Heat Transfer and Transport Systems
Primary Coolant System
Intermediate Coolant System
Auxiliary Cooling Systems
Radioactive Waste Treatment and Disposal
Fuel Handling and Storage Systems
Other Reactor Plant Equipment
Instrumentation and Control
Spart Parts Allowance (1%)
Contingency Allowance (30%)

TURBINE PLANT EQUIPMENT

Turbine-Generators

Main Steam (or other Fluid) System
Heat Rejection System

Other Turbine Plant Equipment
Spare Part Allowance (1%)
Contingency Allowance (20%)

ELECTRIC PLANT EQUIPMENT

Spare Part Allowance (0.5%)
Contingency Allowance (20%)

MISCELLANEQUS PLANT EQUIPMENT

Transportation and Lifting Equipment
Air and Water Service

Other Plant Equipment

Spare Parts Allowance (1%)
Contingency Allowance (20%)

Total Direct Cost

CONSTRUCTION FACILITIES, EQUIP-

MENT AND SERVICES (15%)

ENGINEERING AND CONSTRUCTION
MANAGEMENT SERVICES (15%)

OTHER COSTS (5%)
Total Indirect Cost
TOTAL CAPITAL COST

Refresh
Cycle
2.5

3.41
21.88
10.82
17.02

0.28
11.13
67.03

64.55
19.89
119.38

263.38
253.81

277.41
92.47

1849.40

647.2°
2496.69
Linear Theta Pinch Hybrid Reactor Capital

Blanket

_Type
Account Number

20
21

22

23

24

25

91

92

93

21
21
21
21
21
21

22.

22.

22.
22.
22.
22.
22.
22.
22.

23.
23.
23.
23.
23.
23.

24.
24,

25.
25.
25.
25.
25.

.01
.02
.03
.06
.98
.99

0l

02

03
04
05
06
07
98
99

01
02
03
06
98
99

98
99

01
02
07
98
99

22,
22.
22.
22.
22.
22.
.01
22.

22

22.
22.

01
01

01
01
01

01

02.
02.

.01
.02
01.
.04
.05
.06
.08
.09

03

01
02

(a) June 1978 dollars

LAND AND LAND RIGHTS

STRUCTURES AND SITE FACILITIES

Site Improvements and Facilities
Reactor Building

Turbine Building

Miscellaneous Buildings

Spare Parts Allowance (0.5%)
Contingency Allowance (20%)

REACTOR PLANT EQUIPMENT

Reactor Equipment
Blanket and First Wall
Shield
Magnets
Supplemental Heating
Primary Support and Structure
Reactor Vacuum Systems
Impurity Control
Direct Energy Converter
Main Heat Transfer and Transport Systems
Primary Coolant System
Intermediate Coolant System
Auxiliary Cooling Systems
Radicactive Waste Treatment and Disposal
Fuel Handling and Storage Systems
Other Reactor Plant Equipment
Instrumentation and Control
Spart Parts Allowance (1%)
Contingency Allowance (30%)

TURBINE PLANT EQUIPMENT

Turbine-Generators

Main Steam (or other Fluid) System
Heat Rejection System

Other Turbine Plant Equipment
Spare Part Allowance (1%)
Contingency Allowance (20%)

ELECTRIC PLANT EQUIPMENT

Spare Part Allowance (0.5%)
Contingency Allowance (20%)

MISCELLANEQUS PLANT EQUIPMERWT

Transportation and Lifting Lquipment
Air and Kater Service

Other Plant Equipment

Spare Parts Allowance (1%)
Contingency Allowance (20%)

Total Direct Cost

CONSTRUCTION FACILITIES, EQUIP-
MENT AND SERVICES (15%)

ENGINEERING AND CONSTRUCTION
MANAGEMENT SERVICES (15%)

OTHER COSTS (5%)
Total Indirect Cost

TOTAL CAPITAL COST

A-7

Costs ($106)a

Pu Recvcle/
Once Through

2.5

4.84
96.80
21.78
67.28

0.95
38.14

229.79

34.08
5.55
70.42

60.87

249.82

51.00
332.59
15.52
8.19
245.96
1074.00

84.52
175.80
41.98
106.12
4.08
81.68
494.19-

45.98
0.23
9.19

55.40

38.00

0.38
7.60
45.98

285.28

285.28

95.09

1901.86

665.65
2567.51
Blanket
_Type

Linear Theta Pinch Hybrid Reactor Capital

Account Number

20
21

22

23

24

25

91

92

93

21.
21.
21.
21.
21.
21.

22.

22.

22.
22.
22.
22.
22.
22.
22.

23.
23.
23,
23.
23,
23.

24.
24.

25.
25.
25.
25,
25.

22.
22.
22.
.01

22

22.
22.
22.
22.

22

22.

01
01
01

01
01
01
01

.02.
02.

.01
.02
.03
.04
.05
.06
.08
.09

01
02

(a) June 1978 dollars

LAND AND LAND RIGHTS

STRUCTURES AND SITE FACILITIES

Site Improvements and Facilities
Reactor Building

Turbine Building

Miscellaneous Buildings

Spare Parts Allowance {0.5%)
Contingency Allowance (20%)

REACTOR PLANT EQUIPMENT

Reactor Equipment
Blanket and First Wall
Shield
Magnets
Supplemental Heating
Primary Support and Structure
Reactor Vacuum Systems
Impurity Control
Direct Energy Converter
Main Heat Transfer and Transport Systems
Primary Coolant System
Intermediate Ccolant System
Auxiliary Cooling Systems
Radioactive Waste Treatment and Disposal
Fuel Handling and Storage Systems
Other Reactor Plant Equipment
Instrumentation and Control
Spart Parts Allowance (1%)
Contingency Allowance (30%)

TURBINE PLANT EQUIPMENT

Turbine-Generators

Main Steam (or other Fluid) System
Heat Rejection System

Other Turbine Plant Equipment
Spare Part Allowance {1%)
Contingency Allowance (20%)

ELECTRIC PLANT EQUIPMENT

Spare Part Allowance (0.5%)
Contingency Allowance (20%)

MISCELLANEQUS PLANT EQUIPMENT

Transportation and Lifting Equipment
Air and Water Service

Other Plant Equipment

Spare Parts Allowance (1%)
Contingency Allowance {(20%)

Total Direct Cost

CONSTRUCTION FACILITIES, EQUIP-
MENT AND SERVICES (15%)

ENGINEERING AND CONSTRUCTION
MANAGEMENT SERVICES (15%)

OTHER COSTS (5%)
Total Indirect Cost

TOTAL CAPITAL COST

A-8

Costs ($106)a

Th~Pu

Catalyst

2.

249,

421.
140.

5

82

93
64

2812.87

984.50
3/97 37
Account Number

Linear Theta Pinch Hybrid Reactor Capital

Blanket
_Type

20
21

22

23

24

25

91

92

93

21.
21.
21.
21.
21,
21.

22.

22.
22.
22.
22.
22.
22.
22.
22.

22.
22.

01
01
01
01
01

01
01

02.
02.

.01
.02
.03
.04
.05
01.
.08
.09

06

01
02

(a) June 1978 dollars

LAND AND LAND RIGHTS

STRUCTURES AND SITE FACILITIES

Site Improvements and Facilities
Reactor Building

Turbine Building

Miscellaneous Buildings

Spare Parts Allowance (0.5%)
Contingency Allowance (20%)

REACTOR PLANT EQUIPMENT

Reactor Equipment
Blanket and First Wall
" Shield
Magnets
Suppliemental Heating
Primary Support and Structure
Reactor Vacuum Systems
Impurity Control
Direct Energy Converter
Main Heat Transfer and Transport Systems
Primary Coolant System
Intermediate Coolant System
Auxiliary Cooling Systems
Radioactive Waste Treatment and Disposal
Fuel Handling and Storage Systems
Other Reactor Plant Equipment
Instrumentation and Control
Spart Parts Allowance (1%}
Contingercy Allowance (30%)

TURBINE PLANT EQUIPMENT

Turbine-Generators

Main Steam (or other Fluid) System
Heat Rejection System

Other Turbine Plant Equipment
Spare Part Allowance (1%)
Contingency Allowance (20%)

ELECTRIC PLANT EQUIPMENT

Spare Part Allowance (0.5%)
Contingency Allowance (20%)

MISCELLANEOUS PLANT EQUIPMENT

Transportation and Lifting Equipment
Air and Water Service

Other Plant Equipment

Spare Parts Allowance (1%)
Contingency Allowance (20%)

Total Direct Cost

CONSTRUCTION FACILITIES, EQUIP-
MENT AND SERVICES (15%)

ENGINEERING AND CONSTRUCTION
MANAGEMENT SERVICES (15%)

OTHER COSTS (5%)

Total Indirect Cost

TOTAL CAPITAL COST

Costs ($10°)2

Refresh
Cycle

2.5

4.34
86.80
19.54
60.37

0.86
34.21

206.12

34.17
5.57
70.61

61.04

250.51

45.81
298.75
13.94
7.80
234.12
1022. 32

75.93
157.91
31.27
95.32
3.60
72.09
436.12

41.30

0.21
8.26
49.77

34.13

0.34
6.83
41.30

263.72

263.72
87.91

1758.13

615.35
2373.48
Ignited Tokamak Hybrid Reactor Capital

Account Number

20
21

22

23

24

25

91

92

93

21
21
21
21
21
21

22.

.01
.02
.03
.06
.98
.99

01

.02

.03
.04
.05
.06
.07
.98
.99

.0
.02
.03
.06
.98
.99

.98
.99

.01
.02
.07
.98
.99

22.
22.
22.
.01
22.
22.
22.
22,

22

22,
22.

01
01
01

01
o
0l
0

02.
02.

.0
.02
.03
.04
.05
.06
.08
.09

01
02

{a) June 1978 dollars

LAND AND LAND RIGHTS 2.
STRUCTURES AND SITE FACILITIES 6
Site Improvements and Facilities 72
Reactor Building 30
Turbine Building 91
Miscellaneous Buildings 1.
Spare Parts Allowance ({0.5%) 40,
Contingency Allowance (20%) 247 .
REACTOR PLANT EQUIPMENT
Reactor tquipment
Blanket and First Wall 38.
Shield 30.
Magnets 136.
Supplemental Heating ) 26
Primary Support and Structure 1
Reactor Vacuum Systems 10
Impurity Control 7.
Direct Energy Converter
Main Heat Transfer and Transport Systems
Primary Coolant System 89
Intermediate Coolant System a4
Auxiliary Cooling Systems 31
Radioactive Waste Treatment and Disposal 10,
Fuel Handling and Storage Systems 72
Other Reactor Plant Equipment 48
Instrumentation and Control 7
Spart Parts Allowance (1%} 9
Contingency Allowance {30%) 288
125
TURBINE PLANT EQUIPMENT
Turbine-Generators 132
Main Steam (or other Fluid) System 150
Heat Rejection System 55.
Other Turbine Plant Equipment 42,
Spare Part Allowance (1%) 3.
Contingency Allowance (20%) 76
460.
ELECTRIC PLANT EQUIPMENT 78
Spare Part Allowance (0.5%) 9
Contingency Allowance (20%) 15
o5
MISCELLANEQUS PLANT EQUIPMENT
Transportation and Lifting Equipment b,
Air and Water Service 1.
Other Plant Equipment 5.
Spare Parts Allowance (1%) 0.
Contingency Allowance (20%) 1.
9,
Total Direct Cost .
CONSTRUCTION FACILITIES, EQUIP-
MENT AND SERVICES (T15%) 110.
ENGINEERING AND CONSTRUCTION
MANAGEMENT SERVICES {15%) 310.
OTHER COSTS (5%) 103.

Total Indirect Cost

TOTAL CAPITAL COST

Costs ($106)a

Pu Recycle/

Once Through

2068.48

723.97
2722.45
Ignited Tokamak Hybrid Reactor Capital

Account Number

20
21

22

23

24

25

91

92

93

21.
21.
21.
21.
21.
21.

22.

22.

22.
22.
22.
22.
22.
22.
22.

23.
23.
23.
23.
23.
23.

24.
24.

25.
25.
25,
25.
25.

.01
.02
.03
.04
.05
.06
.08
.09

.01
.02

(a) June 1978 doliars

LAND AND LAND RIGHTS

STRUCTURES AND SITE FACILITIES

Site Improvements and Facilities
Reactor Building

Turbine Building

Miscellaneous Buildings

Spare Parts Allowance (0.5%)
Contingency Aliowance (20%)

REACTOR PLANT EQUIPMENT

Reactor Equipment
Blanket and First Wall
Shield
Magnets
Supplemental Heating
Primary Support and Structure
Reactor Vacuum Systems
Impurity Control
Direct Energy Converter
Main Heat Transfer and Transport Systems
Primary Coolant System
Intermediate Coolant System
Auxiliary Cooling Systems
Radioactive Waste Treatment and Disposal
Fuel Handling and Storage Systems
Other Reactor Plant Equipment
Instrumentation and Control
Spart Parts Allowance (1%)
Contingency Allowance (30%)

TURBINE PLANT EQUIPMENT

Turbine-Generators

Main Steam {or other Fluid) System
Heat Rejection System

Other Turbine Plant Equipment
Spare Part Allowance (1%)
Contingency Allowance (20%)

ELECTRIC PLANT EQUIPMENT

Spare Part Allowance (0.5%)
Contingency Allowance (20%)

MISCELLANEQUS PLANT EQUIPMENT

Transportation and Lifting Equipment
Air and Water Service

Other Plant Equipment

Spare Parts Allowance (1%)
Contingency Allowance (20%)

Total Direct Cost

CONSTRUCTION FACILITIES, EQUIP-
MENT AND SERVICES {15%)

ENGINEERING AND CONSTRUCTION
MANAGEMENT SERVICES (15%)

OTHER COSTS (5%)
Total Indirect Cost

TOTAL CAPITAL COST

A-11

Costs ($106)a

Th-Pu
Catalyst

2.5

36C.83
123.28

2465.56

ovl. 95

3328.51
Ignited Tokamak Hybrid Reactor Capital Costs

Account Number

20
21

22

23

24

25

91

92

93

21
21
21
21
21
21

22.

22.

22

23.
23.
23.
23.
23.
23.

.01
.02
.03
.06
.98
.99

01

02

.03
22.
22.
22.
22.
22.
22.

04
05
06
07
98
99

01
02
03
06
98
99

24.98
24.99

25.01
25.02
25.07
25.98
25.99

22.
22.
22.
22.
22.
22.

22

22.

22.
22.

(a) June 1978 dollars

01
01
01
0
01
0l

01

02.
02.

.01
.02
.03
.04
.05
.06
.01.
.09

08

01
02

LAND AND LAND RIGHTS

STRUCTURES AND SITE FACILITIES

Site Improvements and Facilities
Reactor Building

Turbine Building

Miscellaneous Buildings

Spare Parts Allowance (0.5%)
Contingency Allowance (20%)

REACTOR PLANT EQUIPMENT

Reactor Equipment
Blanket and First Wall
Shield
Magnets
Supplemental Heating
Primary Support and Structure
Reactor Vacuum Systems
Impurity Control
Direct Energy Converter

Main Heat Transfer and Transport Systems

Primary Coolant System
Intermediate Coolant System
Auxiliary Cooling Systems

Radiocactive Waste Treatment and Disposal

Fuel Handling and Storage Systems
Other Reactor Plant Equipment
Instrumentation and Control

Spart Parts Allowance (1%)
Contingency Allowance (30%)

TURBINE PLANT EQUIPMENT

Turuine-Generators

Main Steam (or other Fluid} System
Heat Rejection System

Other Turbine Plant Equipment
Spare Part Allowance (1%)
Contingency Allowance (20%)

ELECTRIC PLANT EQUIPMENT

Spare Part Allowance (0.5%)
Contingency Allowance (20%)

MISCELLANEQUS PLANT EQUIPMENT

Transportation and Lifting Equipment
Air and Water Service

Other Plant Equipment

Spare Parts Allowance (1%)
Contingency Allowance (20%)

Total Direct Cost

CONSTRUCTION FACILITIES, EQUIP-
MENT AND SERVICES (15%)

ENGINEERING AND CONSTRUCTION
MANAGEMENT SERVICES {15%)

OTHER COSTS (5%)
Total Indirect Cost

TOTAL CAPITAL COST

300.89
100. 30

£005.94

702.08

2708.02
APPENDIX B
LEVELIZED ENERGY COST ESTIMATES
LASER ONCE THROUGH
FUEL CYCLE COSTS = NASAR/RYPRID
LFVEL IZED RUSRKRAR EMNERGY COS7T

MILLLS/KWH (197R NOLLARS)
CAPITAL TMVESTMENT CNOST 36,72
HYBRIDN CaPITAl INVESTMEMT <] 38,72
LLwk CAPITAL INVESTMEMNT COSNT .00
OPERATING AND MAINTENANCE COST 6,64
HYBRIND OPERATING AN MAINTENMFNCE CNST hobi
LWR OPFRATING AN{}) MAINTENFNCE COST 0N
FUEL CYCIF ACTIVITY CnSTS R,60
PUR TNTL HLKT uUC « 8N
PUR YRLY HLKT UC 7.13
PiJR INTL RLKT FAR .31
PUR YRLY BLXT FAH 1.31
PUR [MTI. HL®XT 316SS « 36
PR YRLY RLKT 31658S 1.51
PURP INTL TRITItm 05
PUR YPLY NDFUTFRIUA «00
PUR INTL LTTHTUM «30
SHPG HYAD SPNT FUFL YkLY Y
NDISP HYRD SPNT FUFL YKLY 1,69
TOTA} CNST = 51,96
LASFR o RECYCLE
FUEL CYCLE CNSTS = MASAPR/ -YrRRTD
1 FYEL I7ED RUSBAR ENERGY COST
MILLS/KkWH (197R DOLLARS)

- D EDGE G S e uE D e e O e WD O ap oD TR A O W

CAPITAL INVESTMENT CONT 14,9R
HYHRTN CAPTTAL INVFSTMENT COSY An, 72
ILWR CAPTTAL INVESTMEMT CO&T 11.0n1
OPERATING AND MAIMTENMANCE CNSI] 15T
HYRRID OPERATING ANt} #AINIFMENCE €COST 6,64
ILWR NPFRATING AND MaTMTENENCE COST o Th
"FUEL CYCLE ACTIVITY CNSTS 1,75
PUR TNTL ALKT tIC .08
PUR YPLY KLKT UC «3
PUUR [NTL HLKT Fa=R « 09
PUR YRLY HLKT FAH 2N
PUyr INTL RLKT J31ASS « 05
PUIR YRLY HLET 3165% .23
PUR TINTEL TRITIUm e N1
PUR YRPLY DFUTERTiIM + 00
PR TNTH LLITHINMm .08
SHPG HYRN SPNTY FILIFL YkLY o N7
RERPRN YRLY HYH) OuTrRUT b
NISP HYHD REPRO wSTF + 05
PLIR YRLY | Wk Pl Mkijp .15
RIR YRLY (WR Pl fagijw e13
PUR YRIY LWk PU Meyp .11
PUR YRLY 1L WR PU Mki® elN
Pilk YRIY | WR FUFL FAHR +80
PUR YRIY LWR FPRPTL FlifF] 33
REPRG YR Y Lwk OHTRPHT o b be
SHPG | WR SPNT FUKL YRLY .07
NISK | wR RFPRN wSTH « N5
TOTAL COST = 20,40

B-2
L.ASER i = 11 CATAI YST
FUEL CYCIF COSTS = NASAD/YHRATW
LFEYELTZEN BRUSRAR ENERGY €C0OST
MILES/ZKWH (1978 DOLLARS)

CAPITAL THMVESTMEMT CONT 13,15
HYHRIN CAPTTAL IMVESTMENT COST 30,01
I.WR CAPTTAL INVFSTWENT CusT 11,01

OPERATING AND MAINTENAMCE COST 1,79
HYRRID DPFRATING AnD “ATNTENENCE CNST S.64
LWER NPERATING AND MAINTENENCE COST o Th

FUEL CYCIF ACTIVITY CosIS .95
PUR INTL RLKT b2 .01
PUR YRLY HL®T tInZ N0
PUR INTL RL®T THC +N]
PIUR YRLY RLKT TuC +04
PIHR TNTL BLeT P P8
PR INTL M=0 RLKT FAR .07
PR YRLY M= R KT FaH 30
PUR TNTL TRITIUM Y
PIHR YRLY DFUTERTM 00
PUR TMNTL LITRTM 01
SHPG HYHRD SPNT FUFL Ykl Y N3
RFPRO=YR] Y HYRD gnTeyT W19
SHPG HYPN REPRQO 102 YRI Y .03
NISP HYPD REPRCG YSTE YREY 07
Piik YRLY LwkR 11237 reiip N
PR YPLY [WR 11233 MK .12
PLIR YRLY [wR 11233 MriIp .10
PR YRLY LWk LIP3 Mpaip 09
PIIR YRLY LWk FUF| FAR .78
PUIR YRLY L whR FuT( FhLF) «NA
PIIR YPLY L WR FRTL FH+| «NAR
RF2RO YREY Lwk OYTEDT 43
SHOG | WR SPNT FUFI. YL Y «07
NISP | wR REPRO WSTF Y Y NG

TAT4l CNST 17,39

B-3
MIRROR ICF THROHAH
FUEL CYCIE COSTS = NASAB/RYRRID
FFVELIZED RUSHAR ENERGY COST
MILLS/KWH (1978 DOLLARS)

CAPITALL INVESTMENT COST 313,31

HYBRID CAPTYAL INVESTMENT QST 313,31

ILWR CAPITAL INVESTMENT COSTY W00
OPERATING AND MAINTENANCE COST 5%,8%

HYBRID NPERATING AND MAINTENFNCF €COST 55,85

ILWR OPERATING AND MAINTENENCFE COST LN
FUEL CYCIF ACTIVITY CnSTS 35,94

PUR TINTL HLKT tIC 1.67

RUR YRLY R KT LIC R,93

PUR INTL BLKT #aAH 17

PUR YRLY RLKT Fai 5,723

PHR INTL ALKT 31685 1.24

RPIJIR YRLY RLWT 31A%S WS

PUR [NTL TRITIUM .37

PR YRLY DFUTERTHM ' N0

PUR TINTL LITHIUM 1.61

SHRG HYHN SPNT FUFL YRIY 1.86

NDISP HYRN SPNT FYUELL YRLY 7.07
TATAL €OST = 4n%,13

B-4
MIRROR Pl RFCYCLF
FUEL CYCLE COSTS = NASAP/4YHPID
LEVELIZED RUSBAR ENERGY CNSY
MTILLS/KdH (197TA NOLLARS)

CAPITAL INVESTMENT COST °3.83
HYRRIN CAPTITAL INVESTMENT COST 3131.3)
| WR CAPJITAL INVESTMENT CNST 11,01

OPERATING AND MAINTENANCF COST 3,.1n
HYRRID OPERATING AND mMATHTENFNCE COST 65,85
| WR NPERATING AND MATMTENFSNCF CNST o TH

FUEEL CYCILF ACTIVITY COSTS 4,37
PR INTL BLKT UC . 09
PIIR YQLY RLKT uUC + 38
PIIR INTL BREKT Fad « 06
PUR YRLY KLKT FaAH Ph
PIIR INTL RLKT 316SS 07
PR YRLY RLKT 316SS «”R
PUR TNTL TRITIUM NP
PIIR YRILY DEUTERTUM + 00
PR INTL LTITHIUM 06
SHPG HYRD SPNT FUFL YPLY . 0R
REPRQ YRLY HYRD OuTPyT «50
NISP HYRD REPRN WSTE s 06
Ptk YRLY LWR RPU MKl 17
PIIR YRLY LWR PU Mxpp «15
PR YRI Y LWR PH Mx(P e 173
PiIR YRLY ILWR Pl Mi1IP «11
PUR YRLY | WR FIUEL FaR « 91
PR YR Y LWR FRTL FUFL .37
RFPRN YRLY | wR OHTPUT 5N
SHPG | wR SPNT FUFL YRLY .08
DIS® | wR RFPRO wSTF 0k

TOTAL COSTY = 31,74

B-5
ITRROR it = P CAT
'UELL CYCLF CNSTS = NASAP/wYARRLD

CAPTTAL TMVESTMENT (COST

Al YST

HYSBRTID CAPITAL IMVESTMENRT €COST

Iwi CAPTITAl [INVESTMEMT COST
OPERATING AND MAINTENAMCE COS]

HYARTD ODPERATING AND MAINTEMFNMCE COST
ILWR OPERATING AND AATNMTENEMCE COST

FUEL CYCLF ACTIVITY CnSTS
BHR TNTL KLY 102
PYUR YRLY RLKT UDP
PR INTL RLKT THC
P! YRLY HKLKTY TwC
PR INTL RLKT PU
PR INTL M= BI KT FaR
PUR YRLY M=) R{rT FaR
PUR TNTIL. TRITIUM
PR YRLY NEUTERIIM
PUK IMTL L ITHTIM
SHRP(G HYRD SPNT FIHFL YuL Y
RERPRNaYRILY KRYR[O OUTRIIT
SHPG HYPD REPRD 1P YRLY
DISP HYPD REFPRO WSTF YRLY
PiR YRLY [ Wk 11233 MR
PHIR YRLY Lwk 1733 Mxip
PHR YOI Y |LWR UP3] Mxijp
PilR YRLY LwWR U233 Mrllp
Pl YRLY LWR FUIEL FAR
PIIR YRLY LWk FRTI FuUFL
PtiR YRLY | WR FRY|. FufF
RHFPR YRLY LWR QuUTPYT
CHPG | WP SPNT FIFL YRLY
NTSP | WR REPKO wSTE YKL Y

TOoTaL CoOsST

B-6

LEVELTZED RUSBAR ENFRGY COST
MILLS/KWH (1978 NOLI ARS)
16,56
93,17
11.91
184
14,79
o ITA
3.1%
N1
0N
01
« 05
« 31
«NR
« 13
« 01
L00
«0N2
« 03
21
N3
03
15
13
« 11
«09
.8]
N9
+N8
« 45
«07
<05

)

21,55
THETA PINCH ONCE THROUGH
FUEL CYCLE COSTS = NASAP/HYRRID
LFVELIZED RIUSBAR ENFRGY COST
MILLS/KWH (1978 NDOLLARS)
CAPITAlI INVFSTMENT CNST 966,65
HYHRID CAPITAL TINVESTMENT COKT 964K ,6%
LLWR CARPITAI INVESTMENT COST . N0

OPERATING AND MAINTENANCFE COST 172,50
HYRRIN OPERATING AND MATINTFNFENCE CNST 1772.50
LWR OPERATING AND MAINTENEMCE COST 0N

FUEL CYCLF ACTIVITY COSTS 1066,60
PUR INTL HLKT HC 55,79
PHR YOI Y HLKT nC 236,49
PLIR TINTL HLXT FaH «>,77
PijR YRLY BLKT FAR 18R] .61
PUr INTL BLKT 31ASS 4R A7
PR YRLY RILLKT 31ASS 207,28
PUR INTL TRITIUM 1.13
PIR YALY DEUTERIIM L0l
PUR INTL LTITHItim 55,41
SHPG HYHD SPNT FUEL YRLY 49,735
DISP =YRD SPNT FiFL YKLY 187 ,5e

TOTAL COST

B-7

= 2P05,75
THETA PINMCH I RECYTLF
FUEL CYCLE CNSTS = NASAR/LYHRID :
LFYEL T1ZFD RIUISRAR ENERGY COST

MILLLS/KWH (1978 DOLLARS)
CAPTTAL INVESTMENT COST 15,26
HYHRTIN CAPITAL IT~VFETMFST CNAST 966,65
LWR CAPITAL INVESTHMEMT (COST 11,01
OPERATING AND MAINTENANCE COST 1.52
HYBRID QPERATING AND MAINTFHRFNCE CNST 172,50
ILWR OPERATING ANMD MAINTFANENMCE COST o TR
FUFEL. CYCLE ACTIVITY COSTS 2n,68
PUR INTL RLKT uC 5
PUR YRLY RBLKT UC 1.05
PR INTL R KT FAR 19
PR YRLY RALKT FABR oR1
PUYR JINTL RBLKT 31658 P27
PUR YRLY RLKT 3165S 92
PR TNTL TRTITTIUM 01
PitR YRLY DEUTERIHIV « 00
PUR TNMTL LITHIUM «”5
SHPG HYR[D SPNT FilFL YRIY P2
REPRO YRLY HYRD miTeyyt 14,03
DISP HYRD REPRN ASTF o 1R
PUR YRLY LWR Pl Myp .18
PUR YRI Y LWR PU Mx(p .16
PR YRIY LWR Pl Mryp .13
PHR YRLY | WR Pl Mwylip 11
PHIR YRLY (LwkR FIUIFL FAR ok
PHR YRLY ILWR FRTL FLUF| «38
RFPRO YRIY LwR NUTPHT &P
SHPG | wR SPNT FUFL YRLY «0R
DISP { WR RFRRO WSTlF o 0A
TOTAL COST = 7,46

B-8
THETA PIaCH 1 o= PRI CATALYST

FUEL

CYCLE COSTS = NMASHL/RYHILD

CAPITAL TMVESTMENT CNST
HYHRRIN CAPTTAL INVESTMENT 0T

OPERATING

LWR CARPTITAL INVESTMENMT COST
AND MATNTENMAMCE COST

HYBRIDND QOPERATIMNG AN MATHNTENENMCE €COST
LWk OPERATING AND “ATMTEFNENDE COST
FUEL CYCIE ACTIVITY COSIS

PR JMTL RLKT {102

PIIR YRLY HRLKT U102

PUR INMTL HLKT THC

PIIR YKLY RBLKT THC

PUHR INTL HLKT PH

PUR TMT|, M= BRI KT Far
BR YRLY M=) RLKT FAR
PHR INTL TKITIUM

PUHR YRIY DEUTER]UIM

PUR TNTI LLTTHTIUM

SHPG HYRD SENT FHFE|L YWLY
RFPRN=YRL Y HYRD DUTPUIT
SHPG HYPD REPRO HIDP2 YRLY
DISP HYPN RQEPR(O WSTE YRLY
PIIR YRILY (LWR (233 Mk
PR YRLY {whR U233 MR
PIIR YRIY | WR 1233 MRk
Pk YRPLY LWR /P33 vt
PR YRLY | WR FUFL Fawr
PilR YRLY { 4k FRI| Faf
BHR YRLY | WP FRTL Firey
RFORD YRIY | whR OJTO0T
SHEG | wR SPNT FUFL YRLY
DTSk | Wl PFPKDN wSTF Yl Y

YOolTAL COsT

B-9

LFEL TZFD RUSBAR FNERGY €OST
MILLS/kwH (1978 N0 { ARS)
17 .85
6) 32
11,01
1.15%
7.6
o Th
H.la
.Ne
«01
. Né
1R
17
3
- 1Y
«00
N0
+07
10
oh?
10
+0R
015
13
11
. N9
.87
o NW
.NA
45
07
o NA

= 19,15
[3VLTED) Tokava« Tale VM L4
FUE_ ZYSLE Cialh e YaRaB/01 %0
LEVILIZED) R J¥BAR ENERGY [ 1S7
Ml eSS/ (1978 GULLAIYY
Ao e PopegeeVppeee
AP Tl Lvyralezat Jsld 35,18
YLD NABRTTAL Fuwts) B T R 15,15
'.qk‘ CARI Vo {vveSYe b [l IR
JPEIAT NG o - Mafvlinesls (A0 b,¢H
SRS IR R O I R N PR KPS WY G S E | 0,24
VAN O DPERLTING LMY YA 1OTENTLoT £ _ .Y
FUE, Zvliz ACtiwidy 280> g, 42
L) B R A f-'-‘L‘l ney .!1
Dok vAL Y apal gl 1,381
PR TNTL S dr Fo g 1
D n v r omp e B4 JHU
O TNT 9L« 31hR8S 0?3
@ iw ro 0 sl Glmys .q]
PR §S5TL To10jLiA !5
- T I‘QLY "r.dl't“'] [ ."U
PR TNT [ L 21
SHEZ Y n s NP F e te .27
ISP HAvaT 3PNt FOsL TRy 1,04
Tala, st E 'Jb.nl

B-10
[3NITED) 1)l ENEE I BTN
FIEL =¥2LE S50 = wtna2/=1 0],
LeEVi LIZED BJS3aR ENERGY [ ST
MIiLw3Z74%4 (1974 DJLLERD)

\AAZ I X REXR AN T YL LN RN

SAR FAL [versl e o1 2l 1y, b2
syl CERPTTAL pavibal 8~ vyl 45,15
L DaPlLEsL dvwnSfooud {gal BT

IPEIATT NG 8w) wsivieng T fund 1,4
AYRYD Prwaling i callEEaCE (ST b5, 2u
LAR IRESATING ANY M GTESENIT LN WTh

FUS_ SYLLE w20 ividy 8t 3.0
e TETL AL« LT , 4
CANE B A ST A T L BT “Is
Paw { 4T KL &l k79 te
Py ¥y v =L T B4 .1“
BUR TIT HLaTf Bisssw 03
Pim Yo ¥ =L €1 41889 1
Byw ot V1ATT]im "
D i LI T A 4 Tp i gw "0
Bum BNTL LT "ee
Sk -y=) 3Pl FEy tagv "o
REPIT) vy& ¥ =20 )72 41 "au
182 v ) mod o oaile Y
3 viL' L aw B g R 1
e VELY twa B v LS4
QR YIL_Y Qe P M@ .‘e
Pliw vYm v [ ax R 4410 "0
Rk YR ¥ a3 FJE. Fas 'ag
Dl oyt Y e Bl B " 4u
HEPR ) ¥y #2500 "ab
AP, Lem RN L S A S § .'j?
ISP aw g @) 8T TS

rata, L8| z 19,25

B-11
13VITED homavax Joe ¥ st xRt
FUBEa SYSLE Zaaln e vead2/4«1 vi gy
AR el Livistae T [y
SYAS]) LaPJta Lvesat 'ENT L1
A LePIVo javaal ravt Logyat
IFEQATING Ay ~ofNtine Ty PUST
drMl Y o PEwAi ) .y Ay VR ENTR R s LS
BHR OUPERATING ANV sl itpNENTE 8
FOB o CYLLE alvpwile 2 3s0
FION QNI skt g
CONL L NI BERTHY & BRI
Byw NI A1 142
PUw YL Y gl 12
DR TNTL 2 €1 w0y
2 ) T NTL re) A <[ B>
Pk ¥R Y ey W €T Eas
Bw [l tWltjue
Plw Y| Y E TR S
Pan NI (bTmLee
B4P S SvBED IPLT F I i<y
iR ey ATy IR
J4Py “¥0 ) JdEPS, .2 T=LY
ILIP wvit ) AxHe o w1 vy
YW YA LY [ AR QAL
Piw Yur ¥ {aw 1p84 1o ¥
BN Y Y [ Ad J23N vV
D in T"I..Y Lo g4y ra s
Pgw YL Y taw Fitk o k3
Duw YA Y e Bl s
Py vye v | 2@ Fut bore
dEP 1 vy o KRR
L LaH e Sk '_i’:“. vey
I[P om EPa g SNEr Ty
Tara o sl

B-12

LEvE 128D AUS4AR
" leed/Z€Ain (197A

ENESLY 187
DJLLAYS)

12,22
24
11
1,U9
4
2.4
: 15,70
SAPLFaL [rybkntrp vt
4YRRI ) Camitayg
bar LaRl1ay

JIERAT[ NG 4D

Al PERaT vy &

4y n
whu
FJUE ' oY
? 4w
™
P W
By
Pk
P
- TS
- 1]
LT

$P3 Fur, v

PR
-
T
- B
C 3

4Py Kk
713°9

Le

dv Ferg+

Eooaafh @ NADRAILAY ju |

el

Fwee ) TFoT [ 58T

I wug3¥=enl 281
tefNTEaNs O et
VeLsTENENCE L8]

IPEMATING 87) CHUIVIENEUTE O8]

NI
LT |
JNTY,
LR YT
INTL
Y*LV
IvTL
YiLi
1vry

TN
viLv
YLy
'dL'
YA Y

Al

)
- o

VIvIiTy 284
AL 12
T e
RLCT Faa
s Foag
ALLT Y1435
Hpat 15D
1971y
gl Ew ] e
LI'"IJ"

wyad Y { g .A{

| W& P QIR
X P g oW
[ Fh|. B iz
LA Pl owg R
e Fowm_ Fo
YALY ¥ L omrRy
“STE

Thtay, Skt

LEVELIZED RUYHAG ENEQGY 08T
“JowS/7qam (1978 DULLARY)
e eesreevSoppuweeaPappeva

NI
19,548
11,91
f.03
7,14
l’o
FLE)
Y
e
o3
.‘“
04
o0
o1
PR
"3
.h$
L 08
.\,d
2
Jue
e
o T

.79

. 19,57
No. of
Copies

OFFSITE

A. A. Churm

DOE Patent Division
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E. J. Hanrahan
U.S. Department of Energy
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J. Pirro 5 Prof. F. L. Ribe
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