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Nishimura, Masahiro; Fukano, Yoshitaka; Kurisaka, Kenichi; Naruto, Kenichi*
Proceedings of 13th Probabilistic Safety Assessment and Management Conference (PSAM-13) (USB Flash Drive), 12 Pages, 2016/10
Fuel subassemblies of fast breeder reactors (FBRs) are densely arranged and have high power densities. Therefore, the local fault (LF) has been considered as one of the possible initiating events of severe accidents. According to the LF evaluation under the condition of total flow blockage of one sub-channel in the analyses of design basis accident (DBA) for Monju, it was confirmed that the pin failures were limited locally without severe core damage. In addition, local flow blockage (LB) of 66% central planar in the subassembly was investigated as one of the beyond-DBA. However, it became clear that these deterministic analyses were not based on a realistic assumption by experimental studies. Therefore, PRA on LF which was initiated from LB was performed reflecting the state-of-the-art knowledge in this study. As the result, damage propagation from LF caused by LB in Monju can be included in CDF of ATWS or PLOHS in the viewpoint of both probability and consequence.
Zheng, X.; Ishikawa, Jun; Sugiyama, Tomoyuki; Maruyama, Yu
Proceedings of 13th Probabilistic Safety Assessment and Management Conference (PSAM-13) (USB Flash Drive), 10 Pages, 2016/10
Choi, B.; Nishida, Akemi; Itoi, Tatsuya*; Takada, Tsuyoshi*; Furuya, Osamu*; Muta, Hitoshi*; Muramatsu, Ken
Proceedings of 13th Probabilistic Safety Assessment and Management Conference (PSAM-13) (USB Flash Drive), 8 Pages, 2016/10
In this study, we address epistemic uncertainty in structure fragility estimation of nuclear power plants (NPPs). In order to identify and quantify dominant factors in fragility assessment, sensitivity analyses of seismic analysis results are conducted for a target NPP building using a three-dimensional finite element model and a conventional lumped mass model (embedded sway rocking model), and the uncertainty caused by the major factors is then evaluated. The results are used to classify epistemic uncertainty levels in a fragility estimation workflow for NPPs in several stages, and a graded knowledge tree technique, which can be used for future fragility estimations, is proposed.
Takata, Takashi; Azuma, Emiko*
Proceedings of 13th Probabilistic Safety Assessment and Management Conference (PSAM-13) (USB Flash Drive), 10 Pages, 2016/10
A new approach has been developed to assess event sequences under external hazard considering a plant status quantitatively and stochastically so as to take various scenarios into account automatically by applying a Continuous Markov Chain Monte Carlo (CMMC) method coupled with a plant dynamics analysis. In the paper, a tornado and a strong wind are selected as the external hazard to assess the plant safety in a loop type sodium cooled fast reactor (SFR). As a result, it is demonstrated that the various scenarios where the order of the occurrence event and its occurrence time differs from each other can be assessed simultaneously as well as the statistical characteristics of plant parameter such as the coolant temperature. Furthermore, a weight factor is introduced so as to investigate the low failure probability events with a comparative small number of the sampling.
Jang, S.*; Yamaguchi, Akira*; Takata, Takashi
Proceedings of 13th Probabilistic Safety Assessment and Management Conference (PSAM-13) (USB Flash Drive), 11 Pages, 2016/10
The current approach to Level 2 probabilistic risk assessment (PRA) using the conventional event-tree (ET)/fault-tree (FT) methodology requires pre-specifications of event order occurrence and component failure probabilities which may vary significantly in the presence of uncertainties. In the present study, a new methodology is proposed to quantify the level 2 PRA in which the accident progression scenarios are dynamic and interactive with the instantaneous plant state and related phenomena. The accident progression is treated as a continuous Markov process and the transition probabilities are evaluated based on the computation of plant system thermal-hydraulic dynamics. A Monte Carlo method is used to obtain the resultant probability of the radioactive material release scenarios. The methodology is applied to the protected loss of heat sink accident scenario of the level 2 PRA of a generation IV fast reactor.