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Onizawa, Kunio; Masaki, Koichi; Osakabe, Kazuya*; Nishikawa, Hiroyuki*; Katsuyama, Jinya; Nishiyama, Yutaka
Nihon Hozen Gakkai Dai-9-Kai Gakujutsu Koenkai Yoshishu, p.374 - 379, 2012/07
To assure the structural integrity of a reactor pressure vessel (RPV) is known as one of the critical issues to maintain the safe long-term operation of a nuclear power plant. In Japan, the assessment methods for RPV integrity, stipulated in the codes and standards, have been endorsed by the regulatory body. Authors have initiated extensive research on the improvement of structural integrity assessment methods of RPVs. In this paper, we describe some research results obtained from the first-year activity. These include the study on revisiting the technical background of the methods, such as loading conditions, postulated crack definition, the other evaluation methods. In addition, studies on probabilistic methods for the applicability to the current rules and the standardization of the probabilistic analysis methods have been presented.
Masaki, Koichi; Nishikawa, Hiroyuki*; Osakabe, Kazuya*; Katsuyama, Jinya; Onizawa, Kunio
no journal, ,
The structural integrity of a reactor pressure vessel (RPV) should be maintained to ensure the safe long-term operation of a nuclear power plant. The assessment methods for RPV integrity are stipulated in the codes and standards. We compared the assessment methods for RPV integrity under pressurized thermal shock events in Japan and the United States. After reviewing the technical background of the assessment method, we clarified the technical issues based on the latest knowledge in this field. We also identified based on sensitivity analysis that a postulated crack which is provided in the domestic code was more conservative than that of the United States.
Masaki, Koichi; Osakabe, Kazuya*; Nishikawa, Hiroyuki*; Katsuyama, Jinya; Onizawa, Kunio
no journal, ,
The structural integrity of a reactor pressure vessel (RPV) should be maintained to ensure the safe long-term operation of a nuclear power plant. Probabilistic fracture mechanics (PFM) code is required to assess the structural integrity of a RPV during pressurized thermal shock (PTS) events quantitatively. The PASCAL3 for probabilistic analysis have been developed in JAEA. In PTS reevaluation project in the United States, PFM analysis using FAVOR has been conducted to obtain screening criteria for PTS. To consider the future application of the probabilistic method to the code in Japan, we have conducted deterministic and probabilistic analyses using PASCAL3 and evaluated the effects of transient loads on brittle crack initiation and fracture. The results showed that the structural integrity of a RPV could be assessed quantitatively through the evaluation of fracture probabilities.
Katsumata, Genshichiro; Masaki, Koichi*; Osakabe, Kazuya*; Nishikawa, Hiroyuki*; Katsuyama, Jinya; Nishiyama, Yutaka; Onizawa, Kunio
no journal, ,
To assure the structural integrity of a reactor pressure vessel (RPV) is known as one of the critical issues to maintain the safe long-term operation of a nuclear power plant. In Japan, the assessment methods for RPV integrity, stipulated in the codes and standards, have been endorsed by the regulatory body. Authors have initiated extensive research on the improvement of structural integrity assessment methods of RPVs. In this paper, we describe some research results obtained from the second-year activity. These include the study on revisiting the technical background of the methods, such as loading conditions, postulated crack definition, the other evaluation methods. In addition, studies on probabilistic methods for the applicability to the current rules and the standardization of the probabilistic analysis methods have been presented.