@misc{oai:repo.qst.go.jp:00079358,
 author = {Hiwatari, Ryoji and Asakura, Nobuyuki and Uto, Hiroyasu and Tokunaga, Shinsuke and Homma, Yuuki and Miyoshi, Yuuya and Someya, Yoji and Sakamoto, Yoshiteru and Tobita, Kenji and Hiwatari, Ryoji and Asakura, Nobuyuki and Uto, Hiroyasu and Tokunaga, Shinsuke and Homma, Yuuki and Miyoshi, Yuuya and Someya, Yoji and Sakamoto, Yoshiteru and Tobita, Kenji},
 month = {Sep},
 note = {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., 14th International Symposium on Fusion Nuclear Technology},
 title = {Basic Concept and Strategy of Japan’s Fusion Demonstration Plant : JA DEMO},
 year = {2019}
}