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Journal Articles

Development of probabilistic fracture mechanics analysis codes for reactor pressure vessels and piping considering welding residual stress

Onizawa, Kunio; Nishikawa, Hiroyuki; Ito, Hiroto

Proceedings of 7th International Conference on the Integrity of Nuclear Components, p.229 - 239, 2008/07

Probabilistic fracture mechanics (PFM) analysis codes for reactor pressure vessel (RPV) and piping, called as PASCAL series are being developed at JAEA, For an RPV, PASCAL2 code has been developed which can evaluate the conditional probabilities of crack initiation and fracture under transient conditions such as pressurized thermal shock. Recent improvements on the PASCAL2 are related to weld overlay cladding. It is shown that the welding residual stress by cladding affects the fracture probability of RPV to some extent. For piping, considering stress corrosion cracking, PASCAL-SP has been developed. The PASCAL-SP evaluates the probabilities of piping failures such as leakage and break of safety-related piping according to Japanese regulation and rules. Residual stress distribution was determined by parametric FEM analyses after verifying with welding experiments. Effects of welding residual stress distribution as well as inspection accuracy have been studied.

Journal Articles

Recent Japanese research activities on probabilistic fracture mechanics for pressure vessel and piping of nuclear power plant

Kanto, Yasuhiro*; Onizawa, Kunio; Machida, Hideo*; Isobe, Yoshihiro*; Yoshimura, Shinobu*

Proceedings of 7th International Conference on the Integrity of Nuclear Components, p.219 - 228, 2008/07

This paper describes a review of recent Japanese activities on probabilistic fracture mechanics (PFM) analyses. JAEA had sponsored research committees on PFM organized by the Japan Society of Mechanical Engineers and the Japan Welding Engineering Society (JWES) for more than a decade. This work still continues with the same members in JWES. The purpose of the continuous activity is to provide probabilistic approaches in several fields of integrity of reactor components. This paper summarizes some of the latest results of this activity. First topic is evaluation of the JSME rules on Fittness-For-Service from the view of PFM, including reactor pressure vessel with a crack of the allowable size, and effect of sizing accuracy in inspection. The next one is development of new PFM techniques including piping reliability assessment on domestic SCC data and maintenance optimization based on risk and economic models. The last is the international round robin program just starting from 2008.

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