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Journal Articles

Uncertainty quantification of seismic response of reactor building considering different modeling methods

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.

Journal Articles

Evaluation of the effects of differences in building models on the seismic response of a nuclear power plant structure

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

Nihon Jishin Kogakkai Rombunshu (Internet), 20(2), p.2_1 - 2_16, 2020/02

AA2018-0122.pdf:2.15MB

no abstracts in English

Journal Articles

Energy of the $$^{229}$$Th nuclear clock isomer determined by absolute $$gamma$$-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

 Times Cited Count:15 Percentile:86.53(Physics, Multidisciplinary)

Journal Articles

Uncertainty of different modeling methods of NPP building subject to seismic ground motions

Choi, B.; Nishida, Akemi; Shiomi, Tadahiko; Muramatsu, Ken*; Takada, Tsuyoshi*

Proceedings of 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.

Journal Articles

Potential for remote controllable systematization of the method of testing reinforced concrete using guided-wave on rebar

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.

Journal Articles

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.

Journal Articles

Industry development activities and applied laser research activities at Fukui branch of Japan Atomic Energy Agency

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.

Journal Articles

Uncertainty evaluation of seismic response of a nuclear facility using simulated input ground motions

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.

Journal Articles

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

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

Transactions of 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.

Journal Articles

Uncertainty assessment of structural modeling in the seismic response analysis of nuclear facilities

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

Transactions of 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.

Journal Articles

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.

Journal Articles

Probabilistic risk assessment method development for high temperature gas-cooled reactors, 3; Development plan of seismic fragility analysis method

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.

Journal Articles

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.

Journal Articles

Reliability enhancement of seismic risk assessment of NPP as risk management fundamentals; Quantifying epistemic uncertainty in fragility assessment using expert opinions and sensitivity analysis

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.

Journal Articles

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

 Times Cited Count:1 Percentile:12.03(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.

Journal Articles

Seismic response analysis of reactor building and equipment using a 3D-FE model for reliability enhancement of seismic risk assessment of NPP

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

Dai-8-Kai Kozobutsu No Anzensei, Shinraisei Ni Kansuru Kokunai Shimpojiumu (JCOSSAR 2015) Koen Rombunshu (CD-ROM), p.108 - 113, 2015/10

Research and development on next-generation seismic probabilistic risk assessment by using 3D vibration simulators is ongoing to evaluate the seismic safety performance of nuclear plants with high reliability. Most structural PRA uses probabilistic schemes such as the scheme that uses 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. However, these ground motions are not directly linked with their corresponding seismic source characteristics. In this context, the authors propose a methodology based on MCS to reproduce a set of input ground motions to develop an advanced PRA scheme. This paper describes the methodology to reproduce a set of input ground motions briefly and the analytical results of a nuclear plant building and equipment using the set of input ground motions.

Journal Articles

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.

Journal Articles

Reliability enhancement of seismic risk assessment of NPP as risk management fundamentals, 3; Sensitivity analysis for the quantification of epistemic uncertainty on fragility assessment

Nishida, Akemi; Choi, B.; Itoi, Tatsuya*; Takada, Tsuyoshi*; Furuya, Osamu*; Muramatsu, Ken*

Transactions of 23rd International Conference on Structural Mechanics in Reactor Technology (SMiRT-23) (USB Flash Drive), 10 Pages, 2015/08

This study focused on uncertainty-assessment frameworks and the development of relevant software to improve the reliability of seismic probabilistic risk assessment (SPRA) for NPP and promote its further use. This research aimed at contributing to development of implementation guidelines on epistemic uncertainty. Some sensitivity analyses were performed using a three dimensional reactor-building model and a conventional evaluation model by using 3D vibration simulator for NPP of JAEA, and the results were provided to the experts for expert-opinion elicitation. The results of the sensitivity analyses were related to the uncertainty evaluation of the buildings and soil to evaluate the fragility of the equipment. In this paper, those results will be shown in comparison with a conventional evaluation model.

Journal Articles

Seismic damage probability by ground motions consistent with seismic hazard

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

Transactions of 23rd International Conference on Structural Mechanics in Reactor Technology (SMiRT-23) (USB Flash Drive), 10 Pages, 2015/08

In the preceding study, the methodology to generate ground-motion time histories for advanced PRA of NPPs was proposed by Nishida et al.. They are consistent with seismic hazard at reference site, and incorporate uncertainties of seismic-source characteristics. The ground motions utilized in conventional PRA are generated to fit to specified spectra such as UHS, and they are often generated without considering the variation of spectra. Even if it is considered, their inter-period correlations are generally assumed to be 1.0. In this paper, the authors prepared some cases of ground-motions sets. Ground motions are generated to fit to the response spectra calculated from hazard-consistent ground motions. While the target response spectra have the same median for all case, they have different variation and inter-period correlation. The response analyses of general RC structure and PWR building are conducted and the damage frequencies of simplified equipment system are compared.

Journal Articles

Reliability enhancement of seismic risk assessment of NPP as risk management fundamentals, 1; Uncertainty analysis with the SECOM2 code

Muta, Hitoshi*; Muramatsu, Ken*; Furuya, Osamu*; Uchiyama, Tomoaki*; Nishida, Akemi; Takada, Tsuyoshi*

Transactions of 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.

189 (Records 1-20 displayed on this page)