現行の原子炉圧力容器の健全性評価手法に対するPASCAL ver.2を用いた確率論的検討

Probabilistic evaluation of current structural integrity assessment method for reactor pressure vessel using PASCAL ver.2

小坂部 和也; 鬼沢 邦雄 ; 柴田 勝之; 鈴木 雅秀

Osakabe, Kazuya; Onizawa, Kunio; Shibata, Katsuyuki; Suzuki, Masahide

軽水炉構造機器の健全性に関する研究の一環として、平成14年度以降確率論的破壊力学解析コードPASCAL(PFM Analysis of Structural Components in Aging LWR)の改良整備を行っている。このコードは、原子炉圧力容器に加圧熱衝撃(PTS: Pressurized Thermal Shock)等の過渡荷重が発生した場合の破損確率を解析するコードである。現行の健全性評価手法に対するPASCAL ver.2を用いた確率論的検討の例として、日本機械学会の発電用原子力設備規格維持規格に規定されている標準検査にかかわるパラメータの感度解析と、日本電気協会の原子力発電所用機器に対する破壊靭性の確認試験方法に規定されている健全性評価手法に従った決定論解析と確率論解析の結果の相関について述べる。

Fifty-five nuclear power plants of light water reactors (LWRs) in Japan including seven plants which have already reached a 30-year operation have been operated as of September 2006. The probabilistic fracture mechanics (PFM) method has been recently highlighted to rationally incorporate the uncertainties arising from the material properties, defect distribution, inspection quality and so on, unlike the conventional deterministic method. As a part of the materials aging degradation and structural integrity research for LWR components, the PFM analysis code PASCAL has been developed in JAEA. This code evaluates the conditional probabilities of crack initiation and fracture of a reactor pressure vessel (RPV) under transient conditions such as pressurized thermal shock (PTS). Sensitivity analyses for pre-service and in-service inspections have been performed according to JSME S NA1-2004 and the correlation between the results by deterministic analysis according to the JEAC 4206-2004 and probabilistic analysis based on PFM for the integrity of RPV under PTS has also been described.