Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Zheng, X.; Tamaki, Hitoshi; Takahara, Shogo; Sugiyama, Tomoyuki; Maruyama, Yu
Proceedings of Probabilistic Safety Assessment and Management (PSAM16) (Internet), 10 Pages, 2022/09
Kubo, Kotaro; Zheng, X.; Tanaka, Yoichi; Tamaki, Hitoshi; Sugiyama, Tomoyuki; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*
Proceedings of 30th European Safety and Reliability Conference and 15th Probabilistic Safety Assessment and Management Conference (ESREL 2020 and PSAM-15) (Internet), p.2279 - 2286, 2020/11
Probabilistic risk assessment (PRA) is one of the methods used to assess the risks associated with large and complex systems. When the risk of an external event is evaluated using conventional PRA, a particular limitation is the difficulty in considering the timing at which nuclear power plant structures, systems, and components fail. To overcome this limitation, we coupled thermal-hydraulic and external-event simulations using Risk Assessment with Plant Interactive Dynamics (RAPID). Internal flooding was chosen as the representative external event, and a pressurized water reactor plant model was used. Equations based on Bernoulli's theorem were applied to flooding propagation in the turbine building. In the analysis, uncertainties were taken into account, including the flow rate of the flood water source and the failure criteria for the mitigation systems. In terms of recovery action, isolation of the flood water source by the operator and drainage using a pump were modeled based on several assumptions. The results indicate that the isolation action became more effective when combined with drainage.
Zheng, X.; Mandelli, D.*; Alfonsi, A.*; Smith, C.*; Sugiyama, Tomoyuki
Proceedings of 30th European Safety and Reliability Conference and 15th Probabilistic Safety Assessment and Management Conference (ESREL 2020 and PSAM-15) (Internet), p.2176 - 2183, 2020/11
Tanaka, Yoichi; Tamaki, Hitoshi; Zheng, X.; Sugiyama, Tomoyuki
Proceedings of 30th European Safety and Reliability Conference and 15th Probabilistic Safety Assessment and Management Conference (ESREL 2020 and PSAM-15) (Internet), p.2195 - 2201, 2020/11
Narukawa, Takafumi; Yamaguchi, Akira*; Jang, S.*; Amaya, Masaki
Proceedings of 14th International Conference on Probabilistic Safety Assessment and Management (PSAM-14) (USB Flash Drive), 10 Pages, 2018/09
Zheng, X.; Tamaki, Hitoshi; Ishikawa, Jun; Sugiyama, Tomoyuki; Maruyama, Yu
Proceedings of 14th International Conference on Probabilistic Safety Assessment and Management (PSAM-14) (USB Flash Drive), 10 Pages, 2018/09
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.
Silva, K.*; Okamoto, Koji*; Ishiwatari, Yuki*; Takahara, Shogo; Promping, J.*
Proceedings of 12th Probabilistic Safety Assessment and Management Conference (PSAM-12) (USB Flash Drive), 12 Pages, 2014/06
no abstracts in English
Fukano, Yoshitaka; Naruto, Kenichi*; Kurisaka, Kenichi; Nishimura, Masahiro
Proceedings of 12th Probabilistic Safety Assessment and Management Conference (PSAM-12) (USB Flash Drive), 12 Pages, 2014/06
Experimental studies, deterministic and probabilistic and risk assessments (PRAs) on local fault (LF) propagation in sodium cooled fast reactors (SFRs) have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Adventitious fuel pin failures were considered to be the most dominant initiators of LFs in these PRAs because of high frequency of occurrence during reactor operation and possibility of subsequent pin-to-pin failure propagation. Therefore event tree analysis (ETA) on fuel element failure propagation initiated from adventitious fuel pin failure (FEFPA) in Monju was performed in this study based on state-of-the-art knowledge on experimental and analytical studies on FEFPA and reflecting latest operation procedure at emergency in Monju. Probability of adventitious fuel pin failures in SFRs which is the initiating event of this ETA was also updated in this study. It was clarified that FEFPA in Monju was negligible and could be included in core damage fraction of the anticipated transient without scram and protected loss of heat sink in the viewpoint of both probability and consequence.
Yamano, Hidemasa; Nishino, Hiroyuki; Kurisaka, Kenichi; Sakai, Takaaki; Yamamoto, Takahiro*; Ishizuka, Yoshihiro*; Geshi, Nobuo*; Furukawa, Ryuta*; Nanayama, Futoshi*; Takata, Takashi*
Proceedings of 12th Probabilistic Safety Assessment and Management Conference (PSAM-12) (USB Flash Drive), 11 Pages, 2014/06
This paper describes mainly preliminary risk assessment against snow in addition to the project overview. The snow hazard indexes are the annual maximum snow depth and the annual maximum daily snowfall depth. Snow hazard curves for the two indexes were developed using 50 year weather data at the typical sodium-cooled fast reactor site in Japan. In this paper, the snow risk assessment showed less than 10
/reactor-year of core damage frequency. The dominant snow hazard category was the combination of 1-2m/day of snowfall velocity and 0.75-1.0 day of snowfall duration. Sensitivity analyses indicated important human actions, which were the improvement of snow removal velocity and the awareness of snow removal necessity.
Nishida, Akemi; Takada, Tsuyoshi*; Itoi, Tatsuya*; Furuya, Osamu*; Muramatsu, Ken*
Proceedings of 12th Probabilistic Safety Assessment and Management Conference (PSAM-12) (USB Flash Drive), 12 Pages, 2014/06
This study focused on uncertainty-assessment frameworks and utilization of expertise develops methodology for quantification of uncertainty associated with final results from SPRA in the framework of risk management of Nuclear Power Plant (NPP) facilities. This research aimed to contribute to the development of probabilistic models for uncertainty quantification- and software (1); to the aggregation of expert opinions on structure/equipment fragility estimation and development of implementation guidance on epistemic uncertainty (2); and to the study of applicability of newly proposed SPRA models to plant models (3). In particular, we focused on the second goal. There were two different groups of experts used: those in the field of civil engineering, and those in the fields of mechanical engineering. With these groups, we conducted a pilot study on the use of expert-opinion elicitation for identification and quantification of parameters of fragility assessment.
Silva, K.*; Ishiwatari, Yuki*; Takahara, Shogo
Proceedings of Probabilistic Safety Assessment and Management Topical Conference; In light of the Fukushima Dai-ichi Accident (PSAM 2013) (USB Flash Drive), 10 Pages, 2013/04
The "cost per severe accident" was introduced as an index to analyze improvement of accident protection and consequence mitigation strategies. This index consists of various costs and consequences converted into monetary values. The evaluations of these costs and consequences were made based on the systematic consequence analysis using level 2 and 3 PSA. The accident sequences were taken from the results of level 2 seismic PSA. The doses to the public and the number of people affected were calculated using the level 3 PSA code OSCAAR. The data obtained from the open documents on the Fukushima Accident are used as much as possible. Sensitivity analyses are carried out to identify the sensitive assumptions or parameters to the cost per severe accident. Base on the results of the sensitivity analyses, optimized accident protection and consequence mitigation strategies are recommended taking into account the various aspects comprehensively using the cost per severe accident.
Kurisaka, Kenichi; Sakai, Takaaki; Yamano, Hidemasa; Nishino, Hiroyuki; Fujita, Satoshi*; Minagawa, Keisuke*; Yamaguchi, Akira*; Takata, Takashi*
Proceedings of Probabilistic Safety Assessment and Management Topical Conference; In light of the Fukushima Dai-ichi Accident (PSAM 2013) (USB Flash Drive), 10 Pages, 2013/04
This study includes a level-1 PSA related to internal events and a seismic event as a representative external event for Japan sodium-cooled fast reactors with passive safety features and a seismic isolation system. For the internal events, it is necessary to evaluate the passive safety features, of which reliability depends on uncertainties of related physical phenomena. In order to consider the reliability of the passive natural circulation, accident sequences leading to core damage result from natural circulation failure were developed, and the annual frequency of the accident sequences was evaluated for the PSA method study. In respects of seismic event, we developed response analysis method considering the coupling effect of horizontal and vertical shaking on the horizontal seismic isolation characteristics, and developed a fragility evaluation model including the effect of seismic non-linearly for the seismic isolation system.
Yamano, Hidemasa; Sakai, Takaaki; Kurisaka, Kenichi
Proceedings of Probabilistic Safety Assessment and Management Topical Conference; In light of the Fukushima Dai-ichi Accident (PSAM 2013) (USB Flash Drive), 11 Pages, 2013/04
This paper provided a failure probability of the natural circulation heat removal function that is expected to actuate only one direct reactor auxiliary cooling system assuming the loss of two primary reactor auxiliary cooling systems. The probability was obtained in the following steps. At first, a reference case was defined through parametric analyses for a realistic evaluation in the PSA. After selecting key uncertainty parameters according to the PIRT, sensitivity analyses were carried out to develop a response surface. Based on this response surface, Monte Carlo calculations were performed assuming appropriate probability distributions of uncertainty parameters, leading to an exceedance probability in terms of temperature difference from the reference case result. In this study, the failure probability was determined above a safety criterion of the coolant boundary temperature. The failure probability of the reference evaluation was very low.
Okano, Yasushi; Kurisaka, Kenichi; Yamano, Hidemasa; Fujita, Satoshi; Nishino, Hiroyuki; Sakai, Takaaki
Proceedings of Probabilistic Safety Assessment and Management Topical Conference; In light of the Fukushima Dai-ichi Accident (PSAM 2013) (USB Flash Drive), 6 Pages, 2013/04
Liu, Q.; Homma, Toshimitsu
Proceedings of 9th International Probabilistic Safety Assessment and Management Conference (PSAM-9) (CD-ROM), 7 Pages, 2008/00
Fussel-Vesely (FV) and Risk Achievement Worth (RAW) are two commonly used measures in importance ranking of basic events in PSA. Both measures are based on point-estimates of the risk. However, realistic failure characteristics of components are associated with some kinds of uncertainty. The uncertainties of component failures are propagated through the model and bring about the uncertainty of the model risk. Therefore, it is necessary to take uncertainty into consideration when the contribution of a basic event to risk is estimated. By using two fault tree models as examples, the authors calculated the FV and the RAW as well as the uncertainty importance of each basic event. The results show that the uncertainty importance ranking of each basic event does not always agree with the ranking with regard to FV (or RAW). It is argued that uncertainty importance measures provides complementary perspectives of the roles of a basic event in determining the risk.
