Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 96

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Places reached through 11-year academic society collaboration activities

Takada, Tsuyoshi

Nihon Jishin Kogakkai-Shi, (44), p.6 - 11, 2021/10

This report describes the status of nuclear seismic safety assurance in Japan before 2007, followed by the background and main achievements of the activities of the three research committees in collaboration with the Japan Association for Earthquake Engineering (JAEE) and the Atomic Energy Society of Japan (AESJ), and finally a summary of the points that the author considers important as the end of these activities.

JAEA Reports

Investigation and consideration on evaluation of radiation doses to residents in the case of a nuclear emergency

Hashimoto, Makoto; Kinase, Sakae; Munakata, Masahiro; Murayama, Takashi; Takahashi, Masa; Takada, Chie; Okamoto, Akiko; Hayakawa, Tsuyoshi; Sukegawa, Masato; Kume, Nobuhide*; et al.

JAEA-Review 2020-071, 53 Pages, 2021/03

JAEA-Review-2020-071.pdf:2.72MB

In the case of a nuclear accident or a radiological emergency, the Japan Atomic Energy Agency (JAEA), as a designated public corporation assigned in the Disaster Countermeasures Basic Act and the Armed Attack Situation Response Law, undertakes technical supports to the national government and local governments. The JAEA is requested to support to evaluate radiation doses to residents in a nuclear emergency, which is specified in the Basic Disaster Management Plan and the Nuclear Emergency Response Manual. For the dose evaluation, however, its strategy, target, method, structure and so on have not been determined either specifically or in detail. This report describes the results of investigation and consideration discussed in the "Working Group for Radiation Dose Evaluation at a Nuclear Emergency" established within the Nuclear Emergency Assistance and Training Center to discuss technical supports for radiation dose evaluation to residents in the case of a nuclear emergency, and aims at contributing to specific and detailed discussion and activities in the future for the national government and local governments, also within the JAEA.

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

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

Development of seismic counter measures against cliff edges for enhancement of comprehensive safety of nuclear power plants, 10; Avoidance of cliff edge for reactor vessel

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

Proceedings of 25th International Conference on Structural Mechanics in Reactor Technology (SMiRT-25) (USB Flash Drive), 10 Pages, 2019/08

The objective of this study is to assess cliff edge effects, which are greatly important for nuclear power plants. Through assessments of failure probabilities (fragility), this study examined seismic margins of simulated two kinds of thin- and thick-walled reactor vessels by using response waveforms of the reactor building with/without a seismic isolation system obtained by seismic response analyses. The fragility analyses showed that the seismic isolation technology largely reduced the structural response effects nearly twice as much as that of the non-isolated plant. In focusing on uncertainty of response factor of components, the seismic isolation plant has a significant margin compared to the non-isolated plant even if factors from 0.5 to 2.0 are taken into account. This study concluded that the seismic isolation technology is effective to avoid cliff-edge effects.

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

Development of seismic countermeasures against cliff edges for enhancement of comprehensive safety of nuclear power plants; Cliff edges relevant to NPP building system

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

Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 11 Pages, 2018/07

This study identified and quantified possible cliff edge effects through a seismic safety evaluation of a nuclear power plant, based on the concepts of risk and defense in depth. Cliff edges of the both physical and knowledge-based type were considered in this study. We investigated a seismic isolation effect, etc., for physical cliff edges, and the modeling of the target structure, boundary conditions, etc., for knowledge-based cliff edges. Response analysis was performed using a sway-rocking (SR) model and a three-dimensional model of the target building. The seismic isolation effect of the base-isolated building was confirmed by comparison to the results of earthquake-resistant building. In the case of a collision with the retaining wall of the base-isolated building, the level of damage was found to depend on the modeling of the collision condition assumed. On the other hand, the study confirmed the differences between the results from the SR model and the three-dimensional model.

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

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

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

Development of seismic countermeasures against cliff edges for enhancement of comprehensive safety of nuclear power plants, 2; Cliff edges relevant to NPP structure modeling

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

Transactions of 24th International Conference on Structural Mechanics in Reactor Technology (SMiRT-24) (USB Flash Drive), 9 Pages, 2017/08

In this research, the seismic safety of nuclear power plants (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. A methodology is then developed for avoiding these cliff edge effects. The first step was to carry out a preliminary elastic-plastic analysis of the NPP building system. From the analysis, some knowledge was obtained for the modeling factor dependence of cliff edge effects. Next, a preliminary fragility evaluation of the reactor vessel and piping was carried out; it was found that introducing a horizontal seismic isolation system was very effective for avoiding the cliff edge.

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.

96 (Records 1-20 displayed on this page)