Materials Integration by Fission Reactor Irradiation and Essential Basic Studies for Overall Evaluation Presented by N.Yoshida and K.Abe At the J-US Meeting, Feb , 2000, Makuhari / Japan “Development of an Advanced Blanket- Performance under Irradiation and System Integration (tentative)” Overviews of the Japanese Proposal Presented A.Sagara ” Joint meeting of US/J -WS on Power Plant Studies and Advanced Technologies and IEA Task Meeting on Socioeconomic Aspects of Fusion Power UCSD on March , 2000

This project is designed by close cooperation of blanket, materials, tritium, thermofluid and safety research people. Objectives and Key Issues Creation of a design base for “SELF-COOLED LIQUIED BLANKET” and “HIGH-TEMPERATURE GAS-COOLED BLANKET” a) Development of key technologies necessary for the fabrication and operation of the blanket b) Evaluation of irradiation performance of materials systems for the blanket c) Integrated engineering model of blanket system

Structure of the Tasks Task 1: Self-cooling Liquid Blanket subtask1-1: Key Technological R & D subtask1-2: System performance evaluation under neutron irradiation Task 2: High Temperature Gas-cooling Blanket subtask2-1: Key Technological R & D subtask2-2: System performance evaluation under neutron irradiation Task 3: Modeling for System Integration subtask3-1: Fundamental thermofluid experiments and modeling subtask3-2:Multi-scale modeling for advanced blanket systems

SYSTEM AND KEY ISSUES (1) Flibe-Ferritic alloy/Vanadium Alloy System Compatibility, Tritium transportation and recovery, Corrosion protection and tritium barrier by coating, Neutron irradiation effects,Safety handling of Flibe, etc. (2) Litium-Vanadium Alloy System Compatibility, Insulation coating, Neutron irradiation effects,Tritium recovery Task 1: Self-Cooled Liquid Blanket

OBJECTIVE: (1) Estimation of tritium transportation in liquid breeder and coated and uncoated structural materials  Fabrication of TEST POTS of Flibe and Li for tritium experiments (2) Corrosion of materials immersed in the liquid breeder materials (including development of insulator coatings and corrosion protection coatings).  Fabrication of TEST POTS of Flibe and Li for corrosion experiments (3) Safety handling technology for Flibe, involving Be handling issues, through the experiments Subtask1-1: Key Technological R & D( 1/2 )

(4) Mass transfer phenomena under combined interactions between heat, corrosion and tritium in flow condition.  Construction of a Flibe FLOW LOOP 1st step: construction of a Flibe loop system based on the knowledge of corrosion, protective coatings, and tritium transport properties to be derived by the pot tests and other basic thermofluid studies. 2nd step: Study on corrosion behavior in flowing high temperature liquid breeder. 3rd step: combined experiments that introduce tritium into the loop. These experiments will contribute to establishing guideline for the blanket designing. Subtask1-1: Key Technological R & D( 2/2 )

OBJECTIVE 1 Evaluation of corrosion properties of structural materials (coated and uncoated) and coating soundness in liquid breeder environments under neutron irradiation Neutron irradiation in HFIR and/or ATR Li capsule, Flibe capsule, He capsule (for comparison) (candidate fusion reactor structural materials coated for insulation, tritium barrier, and corrosion protection purposes uncoated structural materials, bulk ceramic materials) Post irradiation experiments weight loss, compositional change, interfacial structural change, coating adhesion, bulk mechanical properties, electrical resistivity, tritium permeability Thermal control experiments Subtask1-2; System Performance Evaluation under Neutron Irradiation (1/2)

OBJECTIVE 2 Evaluation of activation and chemical behavior of tritium and other radioisotopes in Flibe Out-of-pile radiation experiments using radioisotope neutron source (Fliqure) From this experiment and the non-irradiation tests in subtask 1-1, system safety issues will be investigated for the Flibe system. Subtask1-2; System Performance Evaluation under Neutron Irradiation (2/2)

SYSTEM AND KEY ISSUES SYSTEM AND KEY ISSUES SiC/SiC-He System hermetic coating and bonding neutron irradiation effects Based on these basic technological developments, thermostructural designing is possible using ceramic composite materials, whose properties are significantly different from those of metallic materials. Task 2: High Temperature Gas-Cooled Blanket

hermetic coating withstanding a high pressure (>10MPa) He gas. hermetic bonding of structural and piping components thermostructural design of components having enhanced thermal-conductive properties. a tritium barrier functioning at high temperature. environmental effects of SiC/SiC composites in high temperature He, including corrosion and oxidation by residual impurities. compatibility with solid breeders such as Li 2 O and multiplier (Be), compatibility as materials systems, such as a combination of SiC/SiC and other metallic low activation materials. Subtask2-1; Key Technological R & D OBJECRIVE OBJECRIVE Study on key technology for fabrication of a high temperature gas (such as He)-cooled blanket with ceramic composite materials(SiC/SiC composites).

Testing materials:Coated and joint specimens simulating a gas- cooled solid breeder blanket loaded into He filled or He circulating irradiation capsules for neutron irradiation. Evaluation:electrical and thermal conductivity, mechanical properties, corrosion of coating under irradiation performance of materials systems and their stability under irradiation in a gas- cooled blanket (taking into account compatibility of structural materials with coolant, solid breeder and multiplier materials) System design guidelines for a high performance blanket Subtask2-2: System Performance Evaluation under Neutron Irradiation OBJECTIVE To evaluate the performance of materials system under synergistic displacement and transmutation -(He and other insoluble elements in SiC) effects.

KEY ISSUES Development of blanket system will be possible by R&D of key technologies and their integration into the blanket system. For promoting this approach, the following studies are important: (1) prediction of blanket component performance. (2) system integration and development/design of materials highly durable in their operation condition. (3) active contribution to reactor system design. Task 3: Modeling for System Integration Development of models for blanket thermofluid and materials performance including some basic experiments

OBJECTIVE: Studies on thermofluid properties in blanket, heat extraction, stability and control of free surface thermofluid and performance in high magnetic field using experiments with a simulant and modeling. (1) KOH43%water Loop(UCLA) Experiments local (microscopic) thermofluid structure (highly relevant to corrosion protection) Accurate channel and free surface experiments including visualization  scaling law for thermofluid properties of Flibe (2) Molten salt HTS Loop(Tohoku Univ.) Experiments thermofluid structure, global heat transfer properties (3) Computational Modeling Subtask3-1: Fundamental Thermofluid Experiments and Modeling Evaluation of feasibility of free surface reactor systems as well as close channel reactor systems from Flibe thermofluid viewpoints.

OBJECTIVE: (1) Model for macroscopic performance of blanket and materials materials quantitative characterization methodology of radiation environment mass transfer models for bulk microstructural and microchemical evolution and for solid interface and surfaces micro-macro correlation model for mechanical properties will be constructed. (2) Engineering models of the blanket and materials mechanical behavior, compatibility, fracture mode evaluation for structural design guideline. (3) Basic Experiments basic experiments such as ion accelerators Subtask3-2: Multi-Scale Modeling for Advanced Blanket Systems