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Muramatsu, Ken; Kubo, Kotaro; Choi, B.; Nishida, Akemi; Takada, Tsuyoshi
Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 10 Pages, 2024/03
no abstracts in English
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
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.
Hirata, Sakiko*; Kusaka, Ryoji; Meiji, Shogo*; Tamekuni, Seita*; Okudera, Kosuke*; Hamada, Shoken*; Sakamoto, Chihiro*; Honda, Takumi*; Matsushita, Kosuke*; Muramatsu, Satoru*; et al.
Inorganic Chemistry, 62(1), p.474 - 486, 2023/01
Times Cited Count：3 Percentile：22.77(Chemistry, Inorganic & Nuclear)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
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.
Choi, B.; Nishida, Akemi; Muramatsu, Ken*; Itoi, Tatsuya*; Takada, Tsuyoshi*
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 5 Pages, 2020/08
After the 2011 Fukushima accident, the seismic regulation for Nuclear Power Plants (NPP) have been strengthened to take countermeasures against accidents beyond design basis conditions. Therefore, the importance of seismic probabilistic risk assessment has drawn much attention. Uncertainty quantification is a very important issue in the fragility assessment for NPP buildings. In this study, the authors focus on the epistemic uncertainty that can be reduced, and aims to clarify the effects due to different modeling methods of NPP buildings on seismic response results. As the first step of this study, the authors compared the effects on seismic response using two kinds of modeling methods. In order to evaluate the effect, seismic response analysis was performed on two types of building models; the three dimensional finite element model and the conventional lumped mass with sway-rocking model. As the input ground motion, the authors adopted 200 types of simulated seismic ground motions generated by fault rupture models with stochastic seismic source characteristics. For the uncertainty quantification, the authors conducted statistical analyses of the effects on seismic response results of two kinds of modeling methods on building response for each input ground motions, and quantitatively evaluated the uncertainty of response considering different modeling methods. In particular, the difference in modeling methods clearly appeared near the openings of the floors and walls. The authors also report on the knowledge about these three-dimensional effects in seismic response analysis.
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photoproduction on the proton near the reaction thresholdIshikawa, 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
Times Cited Count：4 Percentile：39.70(Physics, Nuclear)Choi, B.; Nishida, Akemi; Muramatsu, Ken*; Takada, Tsuyoshi*
Nihon Jishin Kogakkai Rombunshu (Internet), 20(2), p.2_1 - 2_16, 2020/02
no abstracts in English
Th nuclear clock isomer determined by absolute 
-ray energy differenceYamaguchi, 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
Times Cited Count：43 Percentile：89.99(Physics, Multidisciplinary)Choi, B.; Nishida, Akemi; Shiomi, Tadahiko; Muramatsu, Ken*; Takada, Tsuyoshi*
Transactions of the 25th International Conference on Structural Mechanics in Reactor Technology (SMiRT-25) (USB Flash Drive), 8 Pages, 2019/08
In this study, to clarify the influence of the uncertainty of the input seismic ground-motion response of a nuclear power plant (NPP) building, we examined seismic-response analysis results using two different methods of modeling buildings and then compared the results to evaluate effects related to differences between the models. The two methods we used are the three-dimensional (3D) finite-element (FE) model (mainly composed of shell elements) and the conventional sway-rocking (SR) model. Also, using features of the 3D FE model, we analyzed the spatial features of the response results. In this paper, we describe the differences in seismic response obtained by the 3D FE model and the SR model based on simulated input ground motions, and we discuss the influence of the characteristics of the input ground motion on the maximum-response acceleration of the modeled NPP building.
Furusawa, Akinori; Nishimura, Akihiko; Takenaka, Yusuke; Muramatsu, Toshiharu
Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR 2019) (Internet), 4 Pages, 2019/05
The aim of this work presented here is to demonstrate the potential of our method for remote controllable systematization, of testing reinforced concrete based on ultrasonic guided-wave on rebar. In order to investigate how the deteriorated phenomena has the effects on the ultrasonic guided-wave propagating on the rebar, following experiments are conducted. Test pieces used for the experiments are made of bare steel rod and cylindrically pored mortar to be representing the actual reinforced concrete. Irradiating the end face of the rod with nanosecond pulsed laser makes the ultrasonic guided-wave induced, at the other end face, the guided wave signal is measured with ultrasonic receiver. One test piece is with no damage and the other is deteriorated test piece. The deterioration is made by electrolytic corroded method. The guided-wave signal from the deteriorated test piece is measured with respect to each energization time, the change in the waveform is investigated. Analyzing the results from the experiments above, it is found that the deterioration of rebar has remarkable effects on the guided-wave signal. The signal from test piece with no damage has strong peak at both specific frequency and lower region, on the other hand, signals from deteriorated test piece has only at the specific frequency depending on the diameter of the steel rod. Finally, discussion concerning with the experimental results and future perspective for remote controllable systematization of our method is carried out.
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.
Furusawa, Akinori; Takenaka, Yusuke*; Nishimura, Akihiko; Mizutani, Haruki; Muramatsu, Toshiharu
Nihon Hozen Gakkai Dai-14-Kai Gakujutsu Koenkai Yoshishu, p.479 - 480, 2017/08
Industry development activities on applied laser research held at Fukui branch of Japan Atomic Energy Agency are reported. Industry development is inevitable in long-term vision and strategy for developing maintenance technology and establishing decommissioning technology. Fukui branch of JAEA has organized public seminar offering businesses to promote these activities and technology exchange for years. Here some examples offered in the current seminar are introduced concerning with laser technology. Finally, our goal and important point of view are discussed.
Choi, B.; Nishida, Akemi; Muramatsu, Ken*; Takada, Tsuyoshi*
Proceedings of 12th International Conference on Structural Safety & Reliability (ICOSSAR 2017) (USB Flash Drive), p.2206 - 2213, 2017/08
In order to clarify the influence of the difference of modeling method on the variation of the result of seismic response analysis of nuclear facility, seismic response analysis using various simulated input ground motions was carried out and the sensitivity analyses of the variations in seismic response was conducted. In particular, we focused on the maximum acceleration response of reactor building shear walls, the effect of modeling method on response result and the factors of response variation were described and discussed.
Choi, B.; 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
In order to clarify the influence of the modeling method on the result of seismic response analysis of nuclear facility, seismic response analysis using various simulated input ground motions was carried out and the uncertainty of response results were statistically analyzed. In particular, we focused on the difference of the response due to the structural modeling method (a conventional sway-rocking model and 3D FE model), and the relations among the input level, floor position, and response results were described and discussed.
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.
Itoi, Tatsuya*; Nishida, Akemi; Takada, Tsuyoshi*; Hida, Takenori*; Muramatsu, Ken*; Sato, Hiroyuki
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 5 Pages, 2017/04
In this paper, an overview of development plan for seismic PRA methodology for high temperature gas-cooled reactors (HTGRs) is discussed focusing on seismic fragility analysis. The developed seismic fragility analysis has the features as follows: (1) Appropriate treatment of uncertainty in seismic fragility analysis, (2) Utilization of ground motion simulation considering fault rupture process, (3) Utilization of detailed finite element models for seismic fragility analysis. It is also intended that seismic fragility analysis method to be developed is applicable to that of light water reactors.
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.
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.
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.
Nishida, Akemi; Igarashi, Sayaka*; Sakamoto, Shigehiro*; Uchiyama, Yasuo*; Yamamoto, Yu*; Muramatsu, Ken*; Takada, Tsuyoshi*
Nuclear Engineering and Design, 295, p.875 - 886, 2015/12
Times Cited Count：3 Percentile：24.46(Nuclear Science & Technology)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.
