Application of quasi-Monte Carlo and importance sampling to Monte Carlo-based fault tree quantification for seismic probabilistic risk assessment of nuclear power plants

Kubo, Kotaro; Tanaka, Yoichi; Hakuta, Yuto*; Arake, Daisuke*; Uchiyama, Tomoaki*; Muramatsu, Ken

Mechanical Engineering Journal (Internet), 10(4), p.23-00051_1 - 23-00051_17, 2023/08

https://doi.org/10.1299/mej.23-00051

The significance of probabilistic risk assessments (PRAs) of nuclear power plants against external events was re-recognized after the Fukushima Daiichi Nuclear Power Plant accident. Regarding the seismic PRA, handling correlated failures of systems, components, and structures (SSCs) is very important because this type of failure negatively affects the redundancy of accident mitigation systems. The Japan Atomic Energy Research Institute initially developed a fault tree quantification methodology named the direct quantification of fault tree using Monte Carlo simulation (DQFM) to handle SSCs' correlated failures in detail and realistically. This methodology allows quantifying the top event occurrence probability by considering correlated uncertainties related to seismic responses and capacities with Monte Carlo sampling. The usefulness of DQFM has already been demonstrated. However, improving its computational efficiency would allow risk analysts to perform several analyses. Therefore, we applied quasi-Monte Carlo and importance sampling to the DQFM calculation of simplified seismic PRA and examined their effects. Specifically, the conditional core damage probability of a hypothetical pressurized water reactor was analyzed with some assumptions. Applying the quasi-Monte Carlo sampling accelerates the convergence of results at intermediate and high ground motion levels by an order of magnitude over Monte Carlo sampling. The application of importance sampling allows us to obtain a statistically significant result at a low ground motion level, which cannot be obtained through Monte Carlo and quasi-Monte Carlo sampling. These results indicate that these applications provide a notable acceleration of computation and raise the potential for the practical use of DQFM in risk-informed decision-making.

A Scoping study on the use of direct quantification of fault tree using Monte Carlo simulation in seismic probabilistic risk assessments

Kubo, Kotaro; Fujiwara, Keita*; Tanaka, Yoichi; Hakuta, Yuto*; Arake, Daisuke*; Uchiyama, Tomoaki*; Muramatsu, Ken*

Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 8 Pages, 2022/08

https://doi.org/10.1115/ICONE29-88773

After the Fukushima Daiichi Nuclear Power Plant accident, the importance of conducting probabilistic risk assessments (PRAs) of external events, especially seismic activities and tsunamis, was recognized. The Japan Atomic Energy Agency has been developing a computational methodology for seismic PRA, called the direct quantification of fault tree using Monte Carlo simulation (DQFM). When appropriate correlation matrices are available for seismic responses and capacities of components, the DQFM makes it possible to consider the effect of correlated failures of components connected through AND and/or OR gates in fault trees, which is practically difficult when methods using analytical solutions or multidimensional numerical integrations are used to obtain minimal cut set probabilities. The usefulness of DQFM has already been demonstrated. Nevertheless, a reduction of the computational time of DQFM would allow the large number of analyses required in PRAs conducted by regulators and/or operators. We; therefore, performed scoping calculations using three different approaches, namely quasi-Monte Carlo sampling, importance sampling, and parallel computing, to improve calculation efficiency. Quasi-Monte Carlo sampling, importance sampling, and parallel computing were applied when calculating the conditional core damage probability of a simplified PRA model of a pressurized water reactor, using the DQFM method. The results indicated that the quasi-Monte Carlo sampling works well at assumed medium and high ground motion levels, importance sampling is suitable for assumed low ground motion level, and that parallel computing enables practical uncertainty and importance analysis. The combined implementation of these improvements in a PRA code is expected to provide a significant acceleration of computation and offers the prospect of practical use of DQFM in risk-informed decision-making.

scattering length from photoproduction on the proton near the reaction threshold

Ishikawa, Takatsugu*; Fujimura, Hisako*; Fukasawa, Hiroshi*; Hashimoto, Ryo*; He, Q.*; Honda, Yuki*; Hosaka, Atsushi; Iwata, Takahiro*; Kaida, Shun*; Kasagi, Jirota*; et al.

Physical Review C, 101(5), p.052201_1 - 052201_6, 2020/05

https://doi.org/10.1103/PhysRevC.101.052201

Energy of the Th nuclear clock isomer determined by absolute -ray energy difference

Yamaguchi, Atsushi*; Muramatsu, Haruka*; Hayashi, Tasuku*; Yuasa, Naoki*; Nakamura, Keisuke; Takimoto, Misaki; Haba, Hiromitsu*; Konashi, Kenji*; Watanabe, Makoto*; Kikunaga, Hidetoshi*; et al.

Physical Review Letters, 123(22), p.222501_1 - 222501_6, 2019/11

https://doi.org/10.1103/PhysRevLett.123.222501

Epistemic Uncertainty Quantification of Floor Responses for a Nuclear Reactor Building

Choi, B.; Nishida, Akemi; Li, Y.; Muramatsu, Ken*; Takada, Tsuyoshi*

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 9 Pages, 2018/07

After the 2011 Fukushima accident, nuclear power plants are required to take countermeasures against accidents beyond design basis conditions. In seismic probabilistic risk assessment (SPRA), uncertainty can be classified as either aleatory uncertainty, which cannot be reduced, or epistemic uncertainty, which can be reduced with additional knowledge and/or information. To improve the reliability of SPRA, efforts should be made to identify and reduce the epistemic uncertainty caused by the lack of knowledge. In this study, we focused on the difference in seismic response by modeling methods, which is related epistemic uncertainty. We conducted a seismic response analysis with two kinds of modeling methods; a three-dimensional finite-element model and a conventional sway-rocking stick model, by using simulated various input ground motions, which is related to aleatory uncertainty. And then we quantified the seismic floor response results of the various input ground motions of each modeling methods. For the uncertainty quantification related to different modeling methods, we further perform a statistical analysis of the floor response results of the nuclear reactor building. Finally, we discussed how to utilize the results from these calculations for the quantification of uncertainty in fragility analysis for SPRA.

Method for detecting optimal seismic intensity index utilized for ground motion generation in seismic PRA

Igarashi, Sayaka*; Sakamoto, Shigehiro*; Ugata, Takeshi*; Nishida, Akemi; Muramatsu, Ken*; Takada, Tsuyoshi*

Transactions of the 24th International Conference on Structural Mechanics in Reactor Technology (SMiRT-24) (USB Flash Drive), 10 Pages, 2017/08

For the purpose of improving the precision of probabilistic seismic PRA for NPPs, the authors developed the methodology for generating hazard-consistent ground motions based on stochastic fault models which include seismic-source uncertainties by Monte Carlo Simulation. The PRA with HCGMs would require a lot of computer power. The optimization of ground-motions generations is one of the most important subjects for practical application of the PRA method. For optimizing the ground-motions generations, seismic sources for the generations should be selected effectively, and this can be conducted by utilizing optimal seismic index in the hazard analysis. In this study, the method for detecting the optimal seismic intensity index which corresponds with damage probabilities of the target equipment system was developed, and the validity of the proposed method was confirmed for some equipment systems, which has different weak equipment with each other.

Probabilistic risk assessment method development for high temperature gas-cooled reactors, 1; Project overviews

Sato, Hiroyuki; Nishida, Akemi; Ohashi, Hirofumi; Muramatsu, Ken*; Muta, Hitoshi*; Itoi, Tatsuya*; Takada, Tsuyoshi*; Hida, Takenori*; Tanabe, Masayuki*; Yamamoto, Tsuyoshi*; et al.

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04

JAEA, in conjunction with Tokyo City University, The University of Tokyo and JGC Corporation, have started development of a PRA method considering the safety and design features of HTGR. The primary objective of the project is to develop a seismic PRA method which enables to provide a reasonably complete identification of accident scenario including a loss of safety function in passive system, structure and components. In addition, we aim to develop a basis for guidance to implement the PRA. This paper provides the overview of the activities including development of a system analysis method for multiple failures, a component failure data using the operation and maintenance experience in the HTTR, seismic fragility evaluation method, and mechanistic source term evaluation method considering failures in core graphite components and reactor building.

Probabilistic risk assessment method development for high temperature gas-cooled reactors, 2; Development of accident sequence analysis methodology

Matsuda, Kosuke*; Muramatsu, Ken*; Muta, Hitoshi*; Sato, Hiroyuki; Nishida, Akemi; Ohashi, Hirofumi; Itoi, Tatsuya*; Takada, Tsuyoshi*; Hida, Takenori*; Tanabe, Masayuki*; et al.

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04

This paper proposes a set of procedures for accident sequence analysis in seismic PRAs of HTGRs that can consider the unique accident progression characteristics of HTGRs. Main features of our proposed procedure are as follows: (1) Systematic analysis techniques including Master Logic Diagrams are used to ensure reasonable completeness in identification of initiating events and classification of accident sequences, (2) Information on factors that govern the accident progression and source terms are effectively reflected to the construction of event trees for delineation of accident sequences, and (3) Frequency quantification of seismically-initiated accident sequence frequencies that involve multiplepipe ruptures are made with the use of the Direct Quantification of Fault Trees by Monte Carlo (DQFM) method by a computer code SECOM-DQFM.

Hazard-consistent ground motions generated with a stochastic fault-rupture model

Nishida, Akemi; Igarashi, Sayaka*; Sakamoto, Shigehiro*; Uchiyama, Yasuo*; Yamamoto, Yu*; Muramatsu, Ken*; Takada, Tsuyoshi*

Nuclear Engineering and Design, 295, p.875 - 886, 2015/12

https://doi.org/10.1016/j.nucengdes.2015.07.033

Most probabilistic risk assessments (PRA) of structures involve the use of probabilistic schemes such as the scheme using probabilistic seismic hazard and fragility curves. Even when earthquake ground motions are required in Monte Carlo Simulations (MCS), they are generated to fit the specified response spectra, such as uniform hazard spectra at a specified exceedance probability. These ground motions, however, are not directly linked with corresponding seismic source characteristics. In this paper, the authors propose a methodology based on MCS to reproduce a set of input ground motions to develop an advanced PRA scheme that can explain the exceedance probability and sequence of functional loss in a nuclear power plant. These generated motions are consistent with the seismic hazard for the target site and their seismic source characteristics can be recognized in detail.

Study on building function loss evaluated by hazard-consistent ground motions

Igarashi, Sayaka*; Sakamoto, Shigehiro*; Nishida, Akemi; Muramatsu, Ken; Takada, Tsuyoshi*

Dai-8-Kai Kozobutsu No Anzensei, Shinraisei Ni Kansuru Kokunai Shimpojiumu (JCOSSAR 2015) Koen Rombunshu (CD-ROM), p.535 - 541, 2015/10

In this study, building function loss induced by hazard-consistent ground motions (HCGMs), which are consistent with seismic hazard of the reference site and are associated with seismic source characteristics, was evaluated in order to confirm the influence by the variance and/or inter-period correlation of response spectra of ground motions on the resulted damage probabilities of equipment system. Firstly, the statistics values of the response spectra of HCGMs were evaluated, and 3 cases of simulated ground-motions sets are generated so that they fit to the median response spectra of HCGMs. The authors conducted structural response analysis with these ground motions set, and calculated annual damage frequency of equipment system. As a result, it was found that the variance of response spectra was more important factor on damage probability evaluation of systems than inter-period correlation.