著者

川崎 信史 細貝 広視 古橋 一郎 笠原 直人

出版者

一般社団法人日本機械学会

雑誌

年次大会講演論文集 : JSME annual meeting

巻号頁・発行日

vol.2006, no.1, pp.959-960, 2006-09-15

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Thermal transient stress at core support structure of advanced fast reactor was evaluated using thermal hydraulic-structure total analysis method with experimental design. Maximum thermal stress is calculated 15〜18% larger than nominal thermal stress by uncertainty of system parameters. Maximum thermal stress was evaluated 63〜68% larger than nominal thermal stress when predicted by the past deign method, therefore about 40% excessive imaginary stress could be appropriate by thermal hydraulic-structure total analysis.