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 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 = - 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 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 \ 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