Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Shimada, Kazumasa; Sakurahara, Tatsuya*; Farshadmanesh, P.*; Reihani, S.*; Mohagehgh, Z.*
Annals of Nuclear Energy, 197, p.110243_1 - 110243_12, 2024/03
This research improves the realism of Level 3 probabilistic risk assessment (PRA) for nuclear power plants (NPP) to avoid subjective expert judgment when setting evacuation behavior for residents. Therefore, the evacuation speed output by the traffic simulation code MATSim was input to the level 3 PRA code MACCS. Furthermore, to set the priority of the places where road closure is to be considered, a method to evaluate the road closure risk due to the earthquake using the natural disaster risk assessment code HAZUS was developed. Then, the relationship between the evacuation routes and the radiation dose was evaluated for the case study of the Sequoyah NPP adopted in the SOARCA study conducted by the US NRC. As a result, the present study found an evacuation route with low closure risk but causing high radiation dose of residents when it is closed. This showed effectiveness of the proposed Level 3 PRA methodology for supporting decision-makers to enhance evacuation routes.
Futagami, Satoshi; Yamano, Hidemasa; Kurisaka, Kenichi; Ujita, Hiroshi*
Proceedings of PSAM 2023 Topical Conference AI & Risk Analysis for Probabilistic Safety/Security Assessment & Management, 8 Pages, 2023/10
To create an innovation for efficient and effective social implementation of nuclear power plant PRA, automatic construction tool for fault tree architecture and automatic failure judgment tool to construct reliability database are developed by using AI and digitization technology. This paper describes overall development plan of PRA methodology using the AI technology and the progress of automatic FT creation tools development.
Yamaguchi, Yoshihito; Li, Y.
Haikan Gijutsu, 63(12), p.22 - 27, 2021/10
no abstracts in English
Maruyama, Yu; Kita, Toshinobu*; Kuramoto, Takahiro*
Nihon Genshiryoku Gakkai-Shi ATOMO, 62(6), p.328 - 333, 2020/06
no abstracts in English
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.
Okano, Yasushi; Yamano, Hidemasa
Proceedings of International Topical Meeting on Probabilistic Safety Assessment and Analysis (PSA 2015) (USB Flash Drive), p.22 - 31, 2015/04
An external hazard curve of a forest fire is evaluated based on a logic tree. The logic tree consists domains of "forest fire breakout and spread conditions", "weather condition" and "vegetation and topographical conditions". A location nearby a typical nuclear power plant site in Japan was selected, and the frequency of a large forest fire of the location is approximately 1/5 of the average in Japan. Forest fire breakout points were selected considering typical forest fire causes in Japan. The weather conditions are represented by two parameter sets of "temperature-humidity" and "wind direction-wind speed". A number of forest fire simulations were performed to obtain a response surface for a frontal fireline intensity. The hazard curve is therefore evaluated by a Monte Carlo simulation such that the annual exceedance probability is about 1.010 per year for the frontal fireline intensity of 200 kW/m and about 1.310 per year for 300 kW/m.
Tamura, Kazuo*; Iriya, Yoshikazu*
JNC TJ9440 2000-004, 22 Pages, 2000/03
In the probabilistic safety assessment(PSA), the fault tree/event tree technique has been widely used to evaluate accident sequence frequencies. However, event tansition which operators actually face can not be dynamically treated by the conventional technique. Therefore, we have made the dynamic analysis program(DYANA) for event transition for a liquid metal cooled fast breeder reactor. In the previous development, we made basic model for analysis. However, we have a probrem that calculation time is too long. At the current term, we made parallelization of DYANA usig MPI. So we got good performance on WS claster. It performance is close to ideal one.
;
JNC TJ9440 2000-002, 90 Pages, 2000/03
In order to support development of the dynamic reliability analysis program DYANA, analyses were made on the event sequences anticipated under emergency situations using the plant dynamics simulation computer code Super-COPD. In this work 9 sequences were analyzed and integrated into an input file for preparing the functions for DYANA using the analytical model and input data which developed for Super-COPD in the previous work. These sequences could not analyze in the previous work, which were categorized into the PLOHS (Protected Loss of Heat Sink) event.
Ikeda, Takao*; Yoshida, Hideji*; Miki, Takahito*
JNC TJ8400 2000-046, 264 Pages, 2000/02
This report contains discussions about methodology for the selection of parameter values, stochastic approach for the biosphere assessment and biosphere modelling for marine discharge case are described. Regarding the methodology for the selection of parameter values, important aspects for the data selection were discussed, and data selection protocol was developed. Regarding the stochastic approach for the biosphere assessment, it is confirmed that Straightforward Monte Carlo Method and Latin Hypercube Sampling Method are the most adequate based on a literature survey. Then stochastic assessment by using biosphere model that was developed in the second progress report was carried out to check the sensitivity of parameter values. Finally, availability of several kind of assessment models for marine discharge case were discussed. It was confirmed that Multiple Compartment Model was the most applicable. Assessment using Multiple Compartment Model was carried out. The results were compared with those derived by numerical model. As a result, the difference between two models were small enough.
; ;
JNC TJ9440 2000-003, 173 Pages, 1999/03
In FY 1997, a program which evaluates the risk in each phase of the maintenance was developed to support the maintenance planning on an FBR plant. In FY 1998, the GUI (Graphical User Interface) of the program developed in FY 1997 was enhanced for its user-friendlyness including facilitation of data settings and interpretation of the computational results. Specifically, following functions were incorporated : (1)To call associated window displays (editing and reporting display) mutually. (2)To edit the combinations of the systems for their maintenance scheduling. Furthermore, some risk evaluation functions such as the database function for cutsets of accident sequences and tracking function of risk trends were developed and added to its analysis module. A series of test on the programs with the GUI and the analysis module was performed and it was verified that the progam worked correctly.