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Study on structural integrity assessment of reactor pressure vessel based on three-dimensional thermal-hydraulics and structural analyses

三次元熱水力・構造解析による原子炉圧力容器の健全性評価に関する研究

勝山 仁哉; 勝又 源七郎; 鬼沢 邦雄; 渡辺 正*; 西山 裕孝

Katsuyama, Jinya; Katsumata, Genshichiro; Onizawa, Kunio; Watanabe, Tadashi*; Nishiyama, Yutaka

本研究では、加圧熱衝撃(PTS)時の原子炉圧力容器(RPV)の荷重条件をより正確に評価するため、コールドレグ,ダウンカマー及びRPVの三次元モデルを作成し、熱水力解析及び熱構造解析を行った。これらにより、PTSに対する健全性評価で想定する軸方向き裂における周方向応力分布の時刻歴を求めた。この応力分布について、現行のRPVの健全性評価に係る国内規格の評価方法により得られる結果と比較を行った。

For structural integrity assessment on reactor pressure vessel (RPV) of pressurized water reactor during the pressurized thermal shock (PTS) events, temperature of coolant water and heat transfer coefficient between coolant water and RPV are dominant factors. These values can be determined on the basis of thermal-hydraulics (TH) analysis simulating PTS events. Using these values, structural integrity assessment of RPV is performed by thermal-structural analysis, e.g. loading that affects the crack initiation and propagation is evaluated. In this study, we performed the TH and thermal-structural analyses using three-dimensional model of cold-leg, downcomer and RPV to assess loading conditions during the PTS more accurate. We obtained the loading histories at the reactor core region of RPV where a crack is postulated in the structural integrity assessment. Through the comparison between analysis results and current evaluation method, conservatism of current method will be discussed.

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