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Journal Articles

Research and examination of seismic safety evaluation and function maintenance for important equipment in nuclear facilities

Furuya, Osamu*; Fujita, Satoshi*; Muta, Hitoshi*; Otori, Yasuki*; Itoi, Tatsuya*; Okamura, Shigeki*; Minagawa, Keisuke*; Nakamura, Izumi*; Fujimoto, Shigeru*; Otani, Akihito*; et al.

Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 6 Pages, 2021/07

Since the Fukushima accident, with the higher safety requirements of nuclear facilities in Japan, suppliers, manufacturers and academic societies have been actively considering the reconstruction of the safety of nuclear facilities from various perspectives. The Nuclear Regulation Authority has formulated new regulatory standards and is in operation. The new regulatory standards are based on defense in depth, and have significantly raised the levels of natural hazards and have requested to strengthen the countermeasures from the perspective of preventing the simultaneous loss of safety functions due to common factors. Facilities for dealing with specific serious accidents are required to have robustness to ensure functions against earthquakes that exceed the design standards to a certain extent. In addition, since the probabilistic risk assessment (PRA) and the safety margin evaluation are performed to include the range beyond the design assumption in the safety improvement evaluation, it is very important to extent the special knowledge in the strength of important equipment for seismic safety. This paper summarizes the research and examination results of specialized knowledge on the concept of maintaining the functions of important seismic facilities and the damage index to be considered by severe earthquakes. In the other paper, the study on reliability of seismic capacity analysis for important equipment in nuclear facilities will be reported.

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

Development of seismic counter measures against cliff edges for enhancement of comprehensive safety of nuclear power plants, 8; Identification and assessment of cliff edges of NPP structural system

Nishida, Akemi; Choi, B.; Yamano, Hidemasa; Itoi, Tatsuya*; Takada, Tsuyoshi*

Proceedings of 25th International Conference on Structural Mechanics in Reactor Technology (SMiRT-25) (USB Flash Drive), 9 Pages, 2019/08

In this research, the seismic safety of a nuclear power plant (NPP) is treated as a system in which the various cliff edge effects are identified and quantified based on the concepts of risk and defense in depth. An aim of this research is to develop a methodology for avoiding these cliff edge effects. In order to examine how the cliff edge state specified and evaluated in the seismic response analysis of the building system, we investigated the seismic isolation mechanism related to physical cliff edges and the modeling effects of the building system related to knowledge oriented cliff edges. In particular, with regard to knowledge-oriented cliff edges, we quantitatively evaluated the uncertainty within the same floor which is evaluated by a three-dimensional building model and tried to reflect it on the fragility evaluation. This paper presents and discusses these results.

Journal Articles

Thermal-hydraulics technological strategy roadmap 2017; An Approach for continuous safety improvement of LWRs

Itoi, Tatsuya*; Iwaki, Chikako*; Onuki, Akira*; Kito, Kazuaki*; Nakamura, Hideo; Nishida, Akemi; Nishi, Yoshihisa*

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 60(4), p.221 - 225, 2018/04

no abstracts in English

Journal Articles

Engineering applications using probabilistic aftershock hazard analyses; Aftershock hazard map and load combination of aftershocks and tsunamis

Choi, B.; Nishida, Akemi; Itoi, Tatsuya*; Takada, Tsuyoshi*

Geosciences (Internet), 8(1), p.1_1 - 1_22, 2018/01

After the Tohoku earthquake in 2011, we observed that aftershocks tended to occur in a wide region after such a large earthquake. These aftershocks resulted in secondary damage or delayed rescue and recovery activities. However, it is difficult to evaluate the hazards of an aftershock before the main shock due to various uncertainties. For possible great earthquakes, we must make decisions based on such uncertainties, and it is important to quantify the various uncertainties. We previously proposed a probabilistic aftershock occurrence model that is expected to be useful to develop plans for recovery activities after future large earthquakes. In this paper, engineering applications of the proposed approach for probabilistic aftershock hazard analysis are shown for demonstration purposes. One application is to use aftershock hazard maps to plan recovery activities. Another application is to derive load combination equations of the load and resistance factor design (LRFD) considering the simultaneous occurrence of tsunamis and aftershocks for the tsunami-resistant design of tsunami evacuation buildings and nuclear facilities.

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

Formulation of nuclear safety under various induced events, 2; Bases and implementation of countermeasures against external events

Itoi, Tatsuya*; Nakamura, Hideo; Nakanishi, Nobuhiro*

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 58(5), p.318 - 323, 2016/05

no abstracts in English

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

Reliability enhancement of seismic risk assessment of NPP as risk management fundamentals, 2; Quantifying epistemic uncertainty in fragility assessment using expert opinions

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

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

The assessment of seismic safety of nuclear power plant facilities has been performed by identifying and quantifying uncertainties in seismic probabilistic risk assessment (SPRA). For the evaluation process, all uncertainties are classified into either aleatory or epistemic uncertainty for their quantification. In this study, systematic evaluation of the epistemic uncertainty on the seismic fragility of structures and equipment is studied and implemented for a model NPP. There are two expert groups formed in this project: experts in the field of buildings and soil ground (CE experts) and experts in the field of pipe and equipment (ME experts). Along with ample results from relevant sensitivity analyses conducted, elicited opinions are carefully treated and classified into several specific areas and integrated into the form of knowledge tree technique (KTT), all of which can be utilized for future fragility estimation.

Journal Articles

Load combination of aftershocks and tsunami for tsunami-resistant design

Choi, B.; Nishida, Akemi; Itoi, Tatsuya*; Takada, Tsuyoshi*

Proceedings of 12th International Conference on Applications of Statistics and Probability in Civil Engineering (ICASP-12) (USB Flash Drive), 8 Pages, 2015/07

Occurrence of huge tsunami and numerous aftershocks are expected after a gigantic subduction earthquake occurs. Therefore, the important coastal structures (nuclear power plants etc.) must be designed against tsunami as well as ground shaking. When the action effects both from aftershocks and tsunami to the structure occur simultaneously, practically reasonable assessment of load combination from aftershocks and tsunami is needed. In order to treat the load combination problem reasonably, stochastic load combination technique can be used, which requires stochastic modeling of action effects from aftershocks and tsunami. Next, the reliability analysis follows, where load and resistance factors can be obtained under the condition that the conditional target reliability for a limit state function is given. This method is demonstrated at some sites in Japan. Finally, load and resistance factor design format for the tsunami-resistant design is proposed.

Journal Articles

Study on next generation seismic PRA methodology, 2; Quantifying effects of epistemic uncertainty on fragility assessment

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.

Oral presentation

Reliability enhancement of seismic risk assessment of NPP as risk management fundamentals, 1; Program plan and objectives

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.

Oral presentation

Study on reliability enhancement of seismic risk assessment of NPP, 1; Program plan and development of new framework of probabilistic models

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.

Oral presentation

Study on next generation seismic PRA methodology, 1; Program plan and proposal of new mathematical framework

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.

Oral presentation

Sensitivity analysis for quantification of uncertainty in fragility evaluation

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

no journal, , 

This research develops methodology for quantification of uncertainty associated with final results from seismic probabilistic risk assessment (SPRA) in the framework of risk management of NPP facilities. In this research, we take a role of aggregation of expert opinion on structure/equipment fragility estimation and development of implementation guidance on epistemic uncertainty (modeling uncertainty). In this report, a trial evaluation result of quantification of uncertainty using analytical results of seismic response analysis and sensitivity analysis for a model plant are shown.

Oral presentation

Reliability enhancement of seismic risk assessment of NPP as risk management fundamentals, 10; Proposal of a new mathematical framework

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.

Oral presentation

Reliability enhancement of seismic risk assessment of NPP as risk management fundamentals, 12; Study on epistemic uncertainty in fragility evaluation

Takada, Tsuyoshi*; Itoi, Tatsuya*; Muramatsu, Ken*; Nishida, Akemi

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, aggregation of expert opinion on structure/equipment fragility estimation, arrangement of the expert's knowledge and development of implementation guidance on epistemic uncertainty are shown.

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