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Hirouchi, Jun; Takahara, Shogo; Komagamine, Hiroshi*; Watanabe, Masatoshi*; Munakata, Masahiro
Proceedings of Asian Symposium on Risk Assessment and Management 2017 (ASRAM 2017) (USB Flash Drive), 11 Pages, 2017/11
no abstracts in English
Okano, Yasushi; Yamano, Hidemasa
Proceedings of Asian Symposium on Risk Assessment and Management 2017 (ASRAM 2017) (USB Flash Drive), 3 Pages, 2017/11
Forest fire hazard assessment methodologies using a logic tree have been applied for the evaluation of combination hazard curves of a forest fire with lightning as an initiator of a forest fire and with a strong wind being independent from a forest fire. The complex shape of the combinational hazard curve of forest fire and lighting is due to that both lightning and high velocity wind tend to appear under unstable weather conditions, and there is correlation between two hazards. The evaluated combinational hazard curve of forest fire and strong wind for the instantaneous wind velocity over 80 m/s has extremely small frequency in the range below 10
/year.
Mano, Akihiro; Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.
Proceedings of Asian Symposium on Risk Assessment and Management 2017 (ASRAM 2017) (USB Flash Drive), 12 Pages, 2017/11
As a rational inspection methodology, risk informed in-service inspection (RI-ISI) has been widely utilized in in-service inspections of nuclear power plants (NPPs) in several countries. In some of NPPs, an RI-ISI methodology developed by Westinghouse Owners Group (WOG) was applied. As a part of RI-ISI process, extent of examination for important piping segments are determined through the comparisons of leak frequencies with its target value based on the industrial piping leak experiences. The leak frequencies for segments are used as a numerical factor for planning examination based on WOG methodology, and can be evaluated through analyses on the basis of probabilistic fracture mechanics (PFM). In Japan Atomic Energy Agency (JAEA), we have developed a PFM analysis code PASCAL-SP for evaluating leak and rupture probabilities or frequencies of welds in piping of light water reactors taking crack initiation and propagation due to aging degradation mechanisms such as fatigue into consideration. Also, evaluation models of probability of crack detection by non-destructive examination considering the crack type, crack depth and performance of examination team is incorporated in PASCAL-SP. In this study, we investigated the applicability of PASCAL-SP into planning of examination considering the effects of repair methodology, performance of inspection team, and examination time. On the basis of analysis results, it was found that examination plans can be reasonably determined by using PASCAL-SP under several conditions, and it was concluded that the PFM is very effective tools in RI-ISI.
Lu, K.; Miyamoto, Yuhei*; Mano, Akihiro; Katsuyama, Jinya; Li, Y.
Proceedings of Asian Symposium on Risk Assessment and Management 2017 (ASRAM 2017) (USB Flash Drive), 9 Pages, 2017/11
Nowadays, probabilistic fracture mechanics (PFM) has been utilized in several countries as a rational method for structural integrity assessment of important structural components such as reactor pressure vessels (RPVs). In PFM analyses, potential flaws in target components are used to evaluate the failure probability or frequency. Therefore, flaw distributions (i.e., flaw depth and density distributions) in an RPV shall be rationally set as one of the most important influential factors, which are developed during the manufacturing process such as welding. Recently, a Bayesian updating methodology was applied to reflect the inspection results into flaw distributions, and the likelihood functions applicable to the case when flaws are detected in inspections were proposed. However, there may be no flaw indication as the inspection results of some RPVs. The flaw distributions in this situation are important while the corresponding likelihood functions have not been proposed. Therefore, this study proposed likelihood functions to be applicable for both case when flaws are detected and when there is no flaw indication as the inspection results. Based on the proposed likelihood functions, several application examples were given in which flaw distributions were estimated by reflecting the inspection results through Bayesian update. The results indicate that the proposed likelihood functions are useful for estimating the flaw distribution for the case when there is no flaw indication as the inspection results.
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.
Yamano, Hidemasa; Naruto, Kenichi*; Kurisaka, Kenichi; Nishino, Hiroyuki; Okano, Yasushi
Proceedings of Asian Symposium on Risk Assessment and Management 2017 (ASRAM 2017) (USB Flash Drive), 3 Pages, 2017/11
Spent fuels are transferred from a reactor core to a spent fuel pool through an external vessel storage tank (EVST) filled with sodium in sodium-cooled fast reactors in Japan (JSFR). The objective of this study is to identify dominant accident sequences leading to fuel failure by conducting PRA for EVST. The EVST heat removal system in JSFR consists of four independent loops with for primary and secondary ones. Based on the JSFR design information, this study has identified initiating events, event and /fault tree analyses, human reliability analysis, and quantification of accident sequences. Fuel damage frequency of the EVST was evaluated approx. 10
/year in this paper. The main contributor of the fuel damage frequency is the loss of heat removal function of the cooling system. The dominant initiating event was the loss of one circuit of normal heat removal operation.
Zheng, X.; Tamaki, Hitoshi; Shiotsu, Hiroyuki; Sugiyama, Tomoyuki; Maruyama, Yu
Proceedings of Asian Symposium on Risk Assessment and Management 2017 (ASRAM 2017) (USB Flash Drive), 11 Pages, 2017/11
Kimura, Masanori; Hato, Shinji*; Matsubara, Takeshi*; Kanno, Mitsuhiro*; Munakata, Masahiro
Proceedings of Asian Symposium on Risk Assessment and Management 2017 (ASRAM 2017) (USB Flash Drive), 9 Pages, 2017/11
The authors developed a new metabolic model for iodine by combining the respiratory tract model (Publ.66), the gastrointestinal tract model (Publ.30) of the ICRP and the metabolic model for iodine (Johnson's model) in order to evaluate the behavior of radioiodine and stable iodine in the body more realistically. The developed metabolic model indicated that a reduction factor (RF) depends on dosage of stable iodine, timing of the administration of stable iodine, different iodine isotopes (
I - 
I), and age groups. Therefore, the RF was calculated by changing these parameters and then a database of the RF was constructed for the application to the OSCAAR code.
Tamaki, Hitoshi; Yoshida, Kazuo; Abe, Hitoshi; Sugiyama, Tomoyuki; Maruyama, Yu
Proceedings of Asian Symposium on Risk Assessment and Management 2017 (ASRAM 2017) (USB Flash Drive), 9 Pages, 2017/11
An accident of evaporation to dryness caused by boiling of high level radioactive liquid waste (HLLW) is postulated as one of severe accidents caused by the loss of cooling function at the fuel reprocessing plant. This accident can be divided into early boiling stage, late boiling stage and dry-out stage by characteristics of accident evolution. It is important to estimate the amount of fission product (FP) transport between the liquid and gas phases, and the amount of FP deposition on the walls in each stage in order to estimate the release amount of FP to the environment. Various research activities have been carried out for this issue. This paper reviews these activities and presents the recent activities at JAEA for development of simulation code for this type of accident.
Takata, Takashi; Takada, Tsuyoshi*; Narumiya, Yoshiyuki*; Kamiya, Masanobu*; Jimbo, Masakazu*; Muta, Hitoshi*; Hayashi, Kentaro*
no journal, ,
The research committee on "Basic Safety Principles on Earthquake Engineering" was established in Japan Association for Earthquake Engineering (JAEE) to propose the basic safety principles on earthquake engineering for nuclear power plants (NPPs) based on defense in depth and also risk concept. In the committee, three working groups were also established. In working group 1 (WG1), which was parallel established as a subcommittee of "Basic Safety Principles on Earthquake Engineering" in Atomic Energy Society of Japan (AESJ), the contents of the basic safety principles on earthquake engineering has been discussed and the material for discussion has been summarized. This paper summarizes the basic principles and points of contention through the discussion.