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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.
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.
Muta, Hitoshi*; Muramatsu, Ken*; Furuya, Osamu*; Uchiyama, Tomoaki*; Nishida, Akemi; Takada, Tsuyoshi*
Transactions of the 23rd International Conference on Structural Mechanics in Reactor Technology (SMiRT-23) (USB Flash Drive), 10 Pages, 2015/08
Seismic PRA is an effective measure to consider the countermeasures and improvement plans to secure the further safety of nuclear power plants regarding to seismic risk for the earthquake exceeding the reference ground motion. However, the application of the seismic PRA has not been utilized sufficiently so far. One of the reasons is that there is not enough consensus among stakeholders regarding to the evaluation methods and consideration of uncertainty for decision-making. This study proposes the mathematic framework to treat the uncertainty properly related to the evaluation of Core Damage Frequency induced by earthquake, the method to evaluate the fragility utilizing expert knowledge, the probabilistic model to cope with the aleatory uncertainty as well as the development of analyzing code including these considerations for the improvement of the reliability of the method and enhancement of utilization of the products of Seismic PRA.
Liu, Q.; Muramatsu, Ken; Uchiyama, Tomoaki*
JAEA-Data/Code 2008-005, 76 Pages, 2008/03
JAEA-Data-Code-2008-005.pdf:2.03MB
SECOM2-DQFM is developed for seismic reliability analysis of complex engineering systems, such as nuclear power plants. Given that the seismic hazard curve of the location site of a plant, the fault tree and event tree models of this plant were known, if the capacities and responses of components were available, the conditional occurrence probability (or frequency) of the top event of the FT models could be estimated by using SECOM2-DQFM. In addition, the importance of each basic event as well as the occurrence frequency of each accident sequence could also be obtained. Further, the concurrent failure probability of multiple components due to earthquake is considered in SECOM2-DQFM by using the method of Direct Quantification of Fault Tree with Monte Carlo simulation. This report is the English translation of the Japanese version of the user's manual of SECOM2-DQFM.
Liu, Q.; Muramatsu, Ken; Uchiyama, Tomoaki*
JAEA-Data/Code 2008-004, 70 Pages, 2008/03
JAEA-Data-Code-2008-004.pdf:4.54MB
SECOM2-DQFM is developed for seismic reliability analysis of complex engineering systems, such as nuclear power plants. Suppose that the seismic hazard curve of the location site of a plant, the fault tree model and the event tree model of this plant are known. If the capacities and responses of components are available, the conditional occurrence probability (and frequency) of the top event of the fault tree model can be estimated by using SECOM2-DQFM. In addition, the importance of each basic event as well as the occurrence probability (and frequency) of each accident sequence can also be calculated. One feature of SECOM2-DQFM is that the method of Direct Quantification of Fault Tree using Monte Carlo simulation (DQFM) is adopted to evaluate the concurrent failure probability of multiple components. This report is summarized as the user manual of SECOM2-DQFM.
Muramatsu, Ken; Liu, Q.; Uchiyama, Tomoaki*
Journal of Power and Energy Systems (Internet), 2(1), p.122 - 132, 2008/00
A preliminary seismic PSA study was conducted for a two-unit site to examine core damage frequency (CDF) and core damage sequences with consideration of the effect of correlation of component failures. In addition, the effectiveness of cross-connection of emergency diesel generators (EDGs) between adjacent units was also examined. The results showed that calculated CDF depends on the assumptions on component correlations and when the rules for assigning correlation coefficients defined in NUREG-1150 program was adopted, the CDF of a single unit, the CDF of this site and the frequency of simultaneous core damage of both units increased. Besides, it might be possible that simultaneous core damage of both units was caused by different accident sequence pairs. When cross-connection of EDGs between two units was available, the CDF of a single unit, the CDF of this site and the frequency of simultaneous core damage of both units decreased.
Uchiyama, Tomoaki; Oikawa, Tetsukuni; Kondo, Masaaki; Watanabe, Yuichi*; Tamura, Kazuo*
JAERI-Data/Code 2002-011, 205 Pages, 2002/03
JAERI-Data-Code-2002-011.pdf:8.52MB
This report is a user's manual of seismic system reliability analysis code SECOM2 developed at the JAERI for system reliability analysis, which is one of the tasks of seismic probabilistic safety assessment (PSA) of nuclear power plants (NPPs). The SECOM2 code has many functions such as calculation of component and system failure probabilities for given seismic motion levels at the site of an NPP based on the response factor method, calculation of accident sequence frequencies and the core damage frequency (CDF), importance analysis using various indicators, uncertainty analysis, and calculation of the CDF taking into account the effect of the correlations of responses and capacities of components. These analyses require the fault tree (FT) representing the occurrence condition of the system failures and core damage, information about responses and capacities of the components which compose the FT, and seismic hazard curve for the NPP site as input. This report presents calculation method used in the SECOM2 code and how to use those functions in the SECOM2 code.
Oikawa, Tetsukuni; Fukushima, Seiichiro*; Takase, Hidekazu*; Uchiyama, Tomoaki*; Muramatsu, Ken
Transactions of 16th International Conference on Structural Mechanics in Reactor Technology (SMiRT-16) (CD-ROM), 8 Pages, 2001/08
no abstracts in English
Takahara, Shogo; Muramatsu, Ken; Yoshida, Kazuo; Uchiyama, Tomoaki*
no journal, ,
no abstracts in English
Muramatsu, Ken; Liu, Q.; Uchiyama, Tomoaki*
no journal, ,
Aming at proposing effective applications of seismic PSAs for design and risk management of nuclear facilities, a scoping study was conducted for seismic PSA of a multi-unit site to examine core damage frequency (CDF) and potential combinations of core damage sequences and the effectiveness of an accident management measure, namely, the cross connection of emergency diesel generators (EDGs) between adjacent units. Twin hypothetical units (BWR-5 with Mark-II Containment) located in the same site were taken as an example. The system model of these two units was constructed based on the Model plant (JAERI-Research 99-035). The CDF as well as the accident sequences was calculated by using SECOM2, a system reliability analysis code for seismic PSA study. In addition, since the frequency of an accident sequence was affected by the correlation of component failure, sensitivity analysis of the effect of correlation failure was performed. As a result on the accident sequences that cause core damage at this two-unit site, the contribution fraction of the cases of core damage at only one unit was higher than the fraction of cases of simultaneous core damage of both units. Further, it was suggested from the sensitivity analysis of cross-connection of EDGs between these two units that the cross connection would be an effective measure to reduce CDF of this two-unit site since the emergency power for the damaged unit could be supplied from the intact one.
Muramatsu, Ken; Liu, Q.; Uchiyama, Tomoaki*
no journal, ,
A preliminary seismic PSA study was conducted to examine core damage frequency (CDF) and core damage sequences with consideration of the effect of correlation of component failures for a two-unit site. The effectiveness of the cross connection of emergency diesel generators (EDGs) between adjacent units was also examined. The results showed that calculated CDF depends on the assumptions on component correlations and when the rules for assigning correlation coefficients defined in NUREG-1150 program was adopted, the CDF of a single unit, the CDF of this two-unit site and the frequency of simultaneous core damage of both units increased. When cross-connection of EDGs between two units was available, the CDF of a single unit, the CDF of this two-unit site as well as the frequency of simultaneous core damage of both units decreased. In addition, the CDF of this two-unit site was smaller than the CDF of a single unit site.
Muramatsu, Ken*; Takada, Tsuyoshi*; Itoi, Tatsuya*; Nishida, Akemi; Uchiyama, Tomoaki*; Furuya, Osamu*; Fujimoto, Shigeru*; Hirano, Mitsumasa*; Muta, Hitoshi*
no journal, ,
In this research project, study on probabilistic models and treatment of epistemic uncertainty, study on quantification of epistemic uncertainty on fragility assessment and verification of usefulness of proposed method are performed for reliability enhancement of seismic risk assessment. In this report, program plan and objectives are shown as part 1.
Muramatsu, Ken*; Furuya, Osamu*; Fujimoto, Shigeru*; Hirano, Mitsumasa*; Muta, Hitoshi*; Takada, Tsuyoshi*; Itoi, Tatsuya*; Nishida, Akemi; Uchiyama, Tomoaki*
no journal, ,
This project, focusing on uncertainty assessment framework and utilization of expertise, and finally developing relevant computer codes to improve reliability of seismic probabilistic risk assessment (SPRA) and to promote its further use of the SPRA, develops methodology for quantification of uncertainty associated with final results from SPRA in the framework of risk management of NPP facilities. The following scopes are set. (1) Development of framework of probabilistic models for uncertainty quantification and Computer codes (2) Aggregation of expert opinion on structure/equipment fragility estimation and development of implementation guidance on epistemic uncertainty (modeling uncertainty) (3) Study on applicability of SPRA to model plant. In this report, program plan of our research and new probabilistic models and treatment of epistemic uncertainty will be explained as part 1.
Uchiyama, Tomoaki*; Nishida, Akemi; Muramatsu, Ken*
no journal, ,
In this research project, study on probabilistic models and treatment of epistemic uncertainty, study on quantification of epistemic uncertainty on fragility assessment and verification of usefulness of proposed method are performed for reliability enhancement of seismic risk assessment. We are performing examination to treat the epistemic uncertainty in the analysis by using SECOM2-DQFM code developed by JAEA. Development of the program is now performed considered about uncertainty analysis method based on Direct Quantification of Fault Tree using Monte Carlo Simulation (DQFM) method. In this report, the uncertainty analysis method will be explained.
Muramatsu, Ken*; Takada, Tsuyoshi*; Nishida, Akemi; Uchiyama, Tomoaki*; Muta, Hitoshi*; Furuya, Osamu*; Fujimoto, Shigeru*; Itoi, Tatsuya*
no journal, ,
This is the first of a series of papers on a study on next generation seismic probabilistic safety assessment methodology being conducted by the authors as a three-year project "Reliability Enhancement of Seismic Risk Assessment of Nuclear Power Plants as Risk Management Fundamentals", which was started in 2012 and is funded by the Ministry of Education, Culture, Sports, Science & Technology (MEXT) of Japan. This paper gives an overview of the program plan and a proposal of new mathematical framework of S-PRA. We developed a computer code based on SECOM2-DQFM developed by JAEA. The proposed mathematical framework will be built-in it in order to make it possible to estimate the accident sequence occurrence probability and its uncertainty.
Muramatsu, Ken*; Uchiyama, Tomoaki*; Muta, Hitoshi*; Nishida, Akemi
no journal, ,
In this research project, study on probabilistic models and treatment of epistemic uncertainty, study on quantification of epistemic uncertainty on fragility assessment and verification of usefulness of proposed method are performed for reliability enhancement of seismic risk assessment. We are performing examination to treat the epistemic uncertainty in the analysis by using SECOM2-DQFM code developed by JAEA. Development of the program is now performed considered about uncertainty analysis method based on Direct Quantification of Fault Tree using Monte Carlo Simulation (DQFM) method. In this report, parallelization of the SECOM2-DQFM code, improvement to enhance the uncertainty analysis function, and the result of a trial analysis to target the BWR will be explained.
Muramatsu, Ken*; Muta, Hitoshi*; Furuya, Osamu*; Fujimoto, Shigeru*; Takada, Tsuyoshi*; Itoi, Tatsuya*; Nishida, Akemi; Uchiyama, Tomoaki*
no journal, ,
This project, focusing on uncertainty assessment framework and utilization of expertise, and finally developing relevant computer codes to improve reliability of seismic probabilistic risk assessment (SPRA) and to promote its further use of the SPRA, develops methodology for quantification of uncertainty associated with final results from SPRA in the framework of risk management of NPP facilities. The following scopes are set. (1) Development of framework of probabilistic models for uncertainty quantification and Computer codes, (2) Aggregation of expert opinion on structure/equipment fragility estimation and development of implementation guidance on epistemic uncertainty (modeling uncertainty), (3) Study on applicability of SPRA to model plant. In this report, new probabilistic models and treatment of epistemic uncertainty will be explained.
Katai, Yuya*; Watanabe, So; Sano, Yuichi; Abe, Ryoji*; Sakurai, Tomoaki*; Arai, Tsuyoshi*; Uchiyama, Takafumi*; Matsuura, Haruaki*
no journal, ,
no abstracts in English
