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Journal Articles

Sensitivity study on the effects of nondestructive examinations on failure probabilities of reactor pressure vessels

Arai, Kensaku*; Katsuyama, Jinya; Li, Y.

Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 7 Pages, 2018/07

We have developed an analysis code called PASCAL for assessing failure frequencies of RPVs based on probabilistic fracture mechanics. In failure frequency analyses, flaw size distribution in RPVs is one of the most important parameters, and it is determined by considering possible flaws generated during fabrication and the flaw-detection capabilities of nondestructive examinations (NDEs). In this study, the effects of NDEs on failure frequencies of RPVs are evaluated using PASCAL considering probability of detection (POD) in terms of minimum detectable flaw size, the smallest probability of non-detection (PND), and flaw size where POD value reaches the smallest PND.

Journal Articles

Benchmark analyses using probabilistic fracture mechanics analysis codes for reactor pressure vessels

Arai, Kensaku*; Katsuyama, Jinya; Li, Y.

Proceedings of Asian Symposium on Risk Assessment and Management 2017 (ASRAM 2017) (USB Flash Drive), 8 Pages, 2017/11

Probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed to assess structural integrity of aged reactor pressure vessels (RPVs) of light water nuclear power plants by Japan Atomic Energy Agency (JAEA). PASCAL is able to obtain failure frequency such as through-wall cracking frequency (TWCF) of RPVs under several transients including pressurized thermal shock (PTS) event. On the other hand, FAVOR was developed to perform almost the same analysis by Oak Ridge National Laboratory (ORNL) under United States Nuclear Regulatory Commission (USNRC) funding and has been utilized in the US nuclear regulation. To improve the reliability of PFM analysis results of PASCAL, benchmark analyses between PASCAL and FAVOR were performed. This paper provides results of the benchmark analyses using analysis conditions and parameters of the US 3-loop pressurized water reactor (PWR) nuclear power plant. Furthermore, sensitivity analyses relating to differences of analysis models (ex. Embrittlement correlation model) between Japan and the US were also conducted.

Oral presentation

Investigation of effect of dose rate on SCC growth behavior of irradiated materials

Kaji, Yoshiyuki; Miwa, Yukio; Ugachi, Hirokazu; Tsukada, Takashi; Shibata, Akira; Kato, Yoshiaki; Arai, Kensaku*; Nakata, Kiyotomo*

no journal, , 

In order to investigate the effect of dose rate on SCC growth behavior, the SCC growth tests were carried out under simulated boiling water reactor (BWR) water conditions using irradiated materials at different dose rate. It was confirmed that the effect of dose rate on SCC growth rate was considered to be small.

Oral presentation

SCC growth behavior of type 304 stainless steel irradiated under the different dose rates at JMTR

Tsukada, Takashi; Kaji, Yoshiyuki; Ugachi, Hirokazu; Nakano, Junichi; Kondo, Keietsu; Arai, Kensaku*; Nakata, Kiyotomo*

no journal, , 

In order to investigate the effect of neutron dose rate on the mechanical property and stress corrosion cracking (SCC) growth behavior of type 304 stainless steel, the crack growth rate (CGR) test, tensile test and microstructure observation have been conducted after neutron irradiation. The specimens were irradiated up to about 1dpa with different dose rates in the Japan Materials Testing Reactor (JMTR). The radiation hardening increased with the high dose rate condition, and a little difference of radiation-induced segregation at grain boundaries was observed in specimens irradiated by different dose rates. There was no remarkable effect of dose rates on IASCC growth behavior.

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