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; Sagayama, Yutaka
Proceedings of Symposium:Energy&the Environment-Th, 139 Pages, 2003/00
None
;
ANS National Meeting, 0 Pages, 2001/00
None
Proceedings P355-364, p.355 - 364, 2000/00
None
Kurisaka, Kenichi; Kani, Yoshio
PNC TN9410 97-050, 127 Pages, 1997/05
In order to make a fast breeder reactor (FBR) practicable, it is important to make rational categorization of a safety design evaluation event. In this study, for the purpose of providing usefull information into the event categorization, we developed the occurrence frequency data of various abnormal events, presenteda proposal of the event categorization primarily based on the occurrence frequency and examined efficiency of the proposal. We researched and analyzed abnormal event instances of foreign and domestic FBR plants and of domestic light water reactor power plants. On the basis of the analysis, the failure occurrence frequency of the protection system (PS) in the FBR model plants with electric power production of 600MW and 1,000MW was quantified. Making use of results of probabilistic safety assessment study, loss-of-function probability of the mitigation system (MS) was quantified. Some of combinations of PS failure and loss of MS function were selected and their occurrence frequencies were quantified. We examined problems in the current categorization of the safety design evaluation event and presented a new proposal of the event categorization. Merits of the new proposal are to have divided the current category of "accident" into "small accident" and "rare accident", explicitly to treat a multiple failure event not as a collateral analysis condition of supposing "a single failure" but as an event to be evaluated, and to have added a category for the event which is beyond design base, but should be supposed to evaluate depth of the safety design in terms of defense-in-depth. Some candidates ofthe safety design evaluation event were identified and applied to the new proposal. In comparison with the current categorization, we obtained perspective that it was possible to evaluate the safety design more in detail and effectively, especially depth of the safety design such as backup reactor shutdown system, decay heat removal function in a natural ...
Nakai, Ryodai; Hioki, Kazumasa; Sakuma, Takashi; Kani, Yoshio
PNC TN9410 91-335, 62 Pages, 1991/10
Reliabilities of systems and components for LMFBR have been analyzed using CREDO (Centralized Reliability Data Organization) database in order to study the safety design of a large scale FBR based on the operational experiences of LMFBRs. In order to understand the characteristics of FBR-specific components, the comparison of reliabilities between safety and non-safety class components, and the trend of reliabilities on design parameter are evaluated. Based on the component reliability under various operating conditions, the deterministic requirements such as single failure criteria and testing effects are examined using a probabilistic technique. A quantitative technical basis is constructed to study an appropriate safety design policy. Reliabilities of a decay heat removal system are analyzed for various system configurations and success criteria. The dominant contributors to system unreliability such as the importance of support system under a forced circulation and the effectiveness of natural circulation are identified to develop the rational measures for reliability improvement.
*; *
PNC TN9410 90-138, 43 Pages, 1990/09
The incident data on fast breeder reactors (FBRs) in the world have been analyzed and summarized in order to obtain insights into characteristics and trends of those incidents. CREDO (Centralized REliability Data Organization) data and several published documents are referred for this work. Data analysis is performed by two steps. First, the trend analyses of the failure events were performed for the type of system, component, failure cause, corrective action and so on. Next, the data of incidents which occurred after 1979 were selected from all data sources and were analyzed in detail from the viewpoint of safety implication, importance and comprehensiveness of safety evaluation for FBRs. As a result of this analysis, it is concluded that the identified incidents leading to reactor shutdown are enveloped by the events postulated in the safety design/evaluation for domestic FBRs, or they are trivial events that do not affect the safety function of the relevant system.
*; *; Himeno, Yoshiaki; Haga, Kazuo*; Miyake, Osamu; *; *
PNC TN9410 90-119, 58 Pages, 1990/03
With a view to giving reasonable requirements for design of containment features in a large LMFBR, this study discusses the following issues: selection of representative events considered in the safety design/evaluation, consideration of the effect of sodium on FP retention in the "Hypothetical Accident" (site evaluation accident), and evaluation of safety margin against beyond-design-basis events. This report contains some technical documents which are provided to the study group meeting.
; Hioki, Kazumasa*; *; *; *
Proceedings of International Topical Meeting on Probability, Reliability and Safety Assessment (PSA '89), p.810 - 819, 1989/00
None
Aizawa, Kiyoto; ;
Proceedings of International Topical Meeting on Probability, Reliability and Safety Assessment (PSA '89), p.182 - 191, 1989/00
None
; *; *; *; *; *
Nihon Genshiryoku Gakkai-Shi, 28(12), p.1096 - 1128, 1986/12
Times Cited Count：0 Percentile：0.02(Nuclear Science & Technology)*; *; *; *; *; *; *
PNC TN241 85-12, 292 Pages, 1985/03
Computer codes for safety analysis, including systems codes, which are used for the evaluation of LMFBR plants, have been developed along with the progress of the construction and the operation of experimental fast reactor "JOYO". On the other hand, a large number of data has been accumulated at O-arai Engineering Center and other laboratories both in Japan and abroad with the advance of safety R&D on Liquid Metal Fast Breeder Reactor (LMFBR). Various models of analysis have been proposed and many computer codes have been developed which are based on these models. This paper describes the essential part of the models and the functions of the codes, thus developed, which are used for the safety evaluation of LMFBR plant such as the prototype fast breeder reactor "MONJU". Analize codes for sodium leakage accident were modified, so that detailed conditions could be taken into account.
; *; *; *; ; *; ; Aizawa, Kiyoto; *
PNC TN241 81-28, 292 Pages, 1981/11
Computer codes for safety analysis, including systems codes, which are used for the evaluation of LMFBR plants, have been developed along with the progress of the construction and the operation of experimental fast reactor "JOYO". On the other hand, a large number of data has been accumulated at O-arai Engineering Center and other laboratories both in Japan and abroad with the advance of safety R&D on Liquid Metal Fast Breeder Reactor (LMFBR). Various models of analysis have been proposed and many computer codes have been developed which are based on these models. This paper describes the essential part of the models and the functions of the codes, thus developed, which are used for the safety evaluation of LMFBR plant such as the prototype fast breeder reactor "MONJU".
