量研学術機関リポジトリ「QST-Repository」は、国立研究開発法人 量子科学技術研究開発機構に所属する職員等が生み出した学術成果（学会誌発表論文、学会発表、研究開発報告書、特許等）を集積しインターネット上で広く公開するサービスです。 Welcome to QST-Repository where we accumulates and discloses the academic research results(Journal Publications, Conference presentation, Research and Development Report, Patent, etc.) of the members of National Institutes for Quantum Science and Technology.
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Japan’s demonstration plant (JA DEMO) missions are defined as follows: (1) steady and stable power generation beyond several hundred megawatts, (2) availability prospect for commercialization, and (3) overall tritium breeding to fulfill self-sufficiency of fuels. To realize those missions, the basic concept of JA DEMO has been developed based on the feasible technology as applied in the ITER design. First, the major radius R=8.5m of JA DEMO enables to accommodate the CS coil large enough for both pulse and steady-state discharge, to bridge the gap between ITER and DEMO. The fusion power Pf=1.5GW of JA DEMO is determined by the heat-handling capability of the divertor investigated by 2D divertor transport code. Because of neutron irradiation by longer operation than ITER, not only W-mono-block/Cu-alloy-pipe (for the strike-point region) but also W-mono-block/F82H-pipe (for the baffle and dome region) are applied for divertor target. This divertor enables longer operation period than about 1 year, and it contributes to increase the plant availability. The vertical maintenance method is applied for blanket replacement, and the replacement maneuver is originated using firm grip method by both up and down supports of the end-effector. The blanket module of honeycomb structure has pressure tightness against the pressurized water condition to avoid the in-vessel loss of coolant accident. The advanced functional materials (Li2TO3 and Be12Ti) developed in the BA activity are applied to avoid hydrogen generation. This honeycomb blanket module achieves tritium breeding ratio TBR=1.07, and its critical R&D issue is manufacture method by hot isostatic pressing (HIP). As for the toroidal field coil (TFC), the design stress is improved from 667MPa of the ITER condition to 800MPa based on the existing cryogenic steel for high pressure gaseous hydrogen, while TFC follows the radial plate winding method. An operation plan of JA DEMO is also prepared to show the strategy to realize the JA DEMO missions. The operation plan is recently applied to discuss commissioning method, and it reveals requirement of the initial tritium loading. Furthermore, to improve plant availability, the operation plan proposes the high core radiation discharge and the second divertor concept without Cu-alloy-pipe. Those basic concept and operation plan of JA DEMO are the starting point to discuss the shift to the DEMO construction phase in Japan.
会議概要（会議名, 開催地, 会期, 主催者等）
14th International Symposium on Fusion Nuclear Technology
Basic Concept and Strategy of Japan’s Fusion Demonstration Plant : JA DEMO
