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.
Thank you very much for using our website. On the 11th of March 2019, this site was moved from our own network server to the JAIRO Cloud network server. If you previously bookmarked this site, that bookmark will no longer work. We would be grateful if you could bookmark the website again. Thank you very much for your understanding and cooperation.
First wall of blanket in fusion reactor is subject to high surface heat flux as well as high coolant pressure simultaneously. The surface heat flux is 0.3 MW/m2 under ITER operating condition and would increase up to about 1 MW/m2 in a fusion reactor.
The water cooled blanket has been developed and square cooling channel is adopted in the first wall in Japan. The square cooling channel could lead to good heat removal performance due to large effective surface area on heat transfer. However, high peak stress at the corner of the cooling channel might cause fatigue failure under cyclic loading. In contrast the circular cooling channel has advantages in the stress concentration.
Present study focused on the shape of the cooling channel of the first wall. The surface heat flux and the coolant pressure were considered as parameters to clarify the thermal mechanical characteristics of the first wall with square and circular cooling channels. Temperature and stress responses in the first wall were evaluated by using finite element method. Stress intensity increased proportionally to the coolant pressure in the first wall with square cooling channel, but the stress intensity was almost constant in the first wall with circular cooling channel. For the circular cooling channel, tensile stress with the coolant pressure is canceled by the compressive stress with surface heat flux. In addition, the structural integrity of the first walls was evaluated by Design-By-Analysis method in ASME Boiler and Pressure Vessel Code (ASME BPVC). Based on the results allowable surface heat flux and coolant pressure were summarized as a design window of the first wall.
会議概要（会議名, 開催地, 会期, 主催者等）
13th International Symposium on Fusion Nuclear Technology
Thermal mechanical characteristics of blanket first wall with different shape of cooling channel
