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Takata, Takashi; Azuma, Emiko*
Journal of Nuclear Science and Technology, 53(11), p.1749 - 1757, 2016/11
Times Cited Count：5 Percentile：49.14(Nuclear Science & Technology)Margin assessment of a nuclear power plant against external hazards is one of the most important issues after Fukushima Dai-ichi Nuclear Power Plant Accident. In this paper, a new approach has been developed to assess the plant status during external hazards and countermeasures against them in operation quantitatively and stochastically. A Continuous Markov chain Monte Carlo (CMMC) method is applied and coupled with a plant dynamics analysis. In the CMMC method, a subsequence plant status is determined by the latest state (Markov chain) and the status is evaluated from the plant dynamics analysis. A failure or success of safety function of plant component is also evaluated stochastically based on a latest state of plant or hazard. A numerical investigation of plant dynamics analysis against a snow hazard is also carried out in a loop type sodium cooled fast reactor so as to assess the margin against the hazard.
Tamaki, Hitoshi; Hamaguchi, Yoshikane; Yoshida, Kazuo; Muramatsu, Ken
Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10
A PSA procedure for MOX fuel fabrication facilities is being developed at the JAERI. This procedure consists of four steps, which are hazard analysis, accident scenario analysis, frequency evaluation and consequence evaluation. The proposed procedure is characterized by the hazard analysis step. The Hazard analysis step consists of two sub-steps. In the first sub-step, a variety of functions of equipment composing the facility are analyzed to identify potential abnormal events exhaustively. In the second sub-step, these potential events are screened to select abnormal events by using a risk matrix based on the rough estimation of likelihood and maximum unmitigated release of radioactive material. One of the unique technical issues in this research is the estimation of likelihood of criticality event. A method is also proposed as a part of PSA procedure taking into consideration of failure of a computerized control system for MOX powder handling process. The applicability of the PSA procedure was demonstrated through the trial application of it to a model plant of MOX fuel fabrication facility.
Tamaki, Hitoshi; Yoshida, Kazuo; Watanabe, Norio; Muramatsu, Ken
Proceedings of International Topical Meeting on Probabilistic Safety Analysis (PSA '05) (CD-ROM), 11 Pages, 2005/00
A probabilistic safety assessment (PSA) procedure for Mixed Oxide (MOX) fuel fabrication facilities is being developed applicable to nuclear facilities at Japan Atomic Energy Research Institute (JAERI). As part of the PSA procedure, the approach to hazard analysis was established, which consists of two analysis steps: Functional Failure Modes and Effects Analysis (Functional FMEA) and Risk Matrix Analysis. In the Functional FMEA, a variety of functions of equipment composing the facility are analyzed to identify potential abnormal events exhaustively. In the second step, these potential events are screened to select abnormal events as candidate events to be analyzed for frequency and consequence by using two-dimensional matrix based on the rough estimation of likelihood and maximum unmitigated release of radioactive material. The applicability of the hazard analysis approach established was demonstrated through the trial application of the PSA procedure being developed to model plant of MOX fuel fabrication facility.
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.
Furukawa, Kazuo; ; ; ; ; ;
JAERI-M 6073, 116 Pages, 1975/03
no abstracts in English
