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Tamura, Kazuo*; Iriya, Yoshikazu*
JNC TJ9440 2000-004, 22 Pages, 2000/03
JNC-TJ9440-2000-004.pdf:2.35MB
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.
Nomura, Yasushi; Naito, Yoshitaka
Nuclear Technology, 121(1), p.3 - 12, 1998/01
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)no abstracts in English
Hioki, Kazumasa
PNC TN9410 94-188, 160 Pages, 1994/05
The Systems Analysis Section has been performing a probabilistic Safety Assessment (PSA) on a large fast breeder reactor (FBR) since JFY 1992. The objective of the study is to apply the PSA method to a plant in a conceptual design stage, develop system models, perform quantitative analyses and systematic evaluation, supply valuable insights to enhance reliability and safety, and reflect them to the basic design. The plant analyzed is a 600MWe class large FBR designed by the Plant Engineering Section in the "Large FBR design study" that has been performed since JFY 1990. The failure probability of the Decay Heat Removal System (DHRS) can be reduced approximately two orders if the Water Steam System (WSS) can remove the decay heat for the first 24 hours. The frequency of PLOHS, however, is not reduced to less than one third because the WSS cannot be used for some initiating events and the PLOHS frequency is dominated by the failure probability of DHRS without the WSS. The failure probability of DHRS is dominated by the common cause failures (CCFs) of vanes, dampers and valves around the air-coolers in the Auxiliary Cooling System (ACS). Therefore it is most important to eliminate the CCFs. Assuming that the CCFs have been eliminated by diversifying the components, the frequencies of PLOHS were evaluated. An analysis has shown that if the WSS can remove the decay heat alone, the PLOHS frequency is reduced approximately two orders. In this case the PLOHS frequency is dominated by the failure probability of the DHRS right after the reactor shutdown. The most effective way to reduce the PLOHS frequency is to increasc the redundancy of the DHRS for the first few hours after reactor shutdown. It is known through the experience of preceding plants that the success criteria can be relaxed to one loop natural circulation instead of forced circulation in the best estimate evaluation. It was shown that under such condition, the PLOHS frequency can be as low as 10
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Nomura, Yasushi
Journal of Nuclear Science and Technology, 29(8), p.813 - 823, 1992/08
no abstracts in English
Watanabe, Norio; *
JAERI-M 92-089, 49 Pages, 1992/06
no abstracts in English
Watanabe, Norio; Kondo, Masaaki; Abe, Kiyoharu
Use of Probabilistic Safety Assessment for Operational Safety:PSA 91, p.61 - 72, 1992/00
no abstracts in English
Nomura, Yasushi
JAERI-M 91-160, 176 Pages, 1991/10
no abstracts in English
PNC TN9410 91-286, 117 Pages, 1991/08
A conventional type of RSS in a large scale FBR was designed and its unavailability was analyzed with fault-tree. Reliability of logic circuits of the reaetor protection system is relatively high when compared to that of the control rod insertion. Contributing factors to the unavailabity are multiple failures of detection systems, and failure to insert rods such as failure to deratch or rod jamming. Then the new concept of control rod release mechanism was introduced in the RSS design. The thermal-hydraulic characteristics of the mechanism was analyzed using computer codes SSC-L and AQUA. Further, qualitative analysis of the common cause failure for the RSS was tried with the generic cause approach. The reactor protection systems of the backup RSS are diversified by the self actuated control rod release mechanism. With such a mechanism, the number of common cause factors were decreased for postulated LOF event.
Nomura, Yasushi
Proc. of the CSNI Specialist Meeting on Safety and Risk Assessment in Fuel Cycle Facilities, p.414 - 426, 1991/00
no abstracts in English
Iguchi, Yukihiro
PNC TN3410 88-007, 100 Pages, 1988/04
In "Fugen", we started the project which evaluates the importance of the components of the plant in 1985, in order to improve the reliability of the plant effectively. The data base of the evaluation are mostly based on the disclosed databases of other reports and partly on the Maintenance Management System (MMS) data base. As a method of the evaluation, Fault Tree Analysis (FTA) is adopted. In 1987, we complete the fault tree of the plant shut-down, and coded all the data as FT data base. Moreover, we collected new data for future use such as maintenance items use and coded them also. In order to analyze the FT data base, we introduce SETS and FTD codes which are used in PNC often. And we developed a transforming program to apply FT database to SEST. This program can manipulate the trains of the components and their logic combination (e.g. 1 out of 2 twice) easily. And we also modified FTD so that it can plot out Japanese characters, because the event names of FT data base are written in Japanese. In future, these program will be used and improved as ATR Maintenance Instruction System (AMIS).
