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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.
Sakuma, Takashi
PNC TN9410 91-324, 79 Pages, 1991/09
System for Human Error and Reliability Investigation SHERI which is mainly based on THERP method has been developed as a part of the probabilistic safety assessment (PSA) of a liquid metal fast breeder reactor. THERP was developed by A. D. Swain as a method to perform probabilistic evaluation of human error in complicated systems such as nuclear power plant. This method has been widely used in PSAs such as WASH-1400, Zion, Indian Point, IREP, and ASEP. However THERP is applied using a handbook of huge pages, and it is only possible to perform analysis correctly provided that he is well acquainted with the methodology. So THERP has been applied actually by an expert. SHERI is a system which loads knowledge about human reliability analysis (HRA) on computor program, and support the human reliability analyst. It is aimed to perform human reliability analysis consistently, even if the analyst is not an expert of HRA. It is expected to decrease misunderstanding in the analysis and to reduce workload of the analyst.
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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.
PNC TN2410 89-011, 104 Pages, 1990/08
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 domestic 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.
; Hioki, Kazumasa*; ; ;
Proceedings of International Topical Meeting on Probability, Reliability and Safety Assessment (PSA '89), p.810 - 819, 1989/00
None
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PNC TN941 80-211, 129 Pages, 1980/11
This report describes the results of the thermal power calibration test (PT-11) that was planned and performed as part of the 75MW power-up testing of the Experimental Fast Reactor "JOYO". The purpose of this test is to calibrate the Power Range Monitors (PRM) by measuring the reactor thermal power at several levels from low power through the rated power of 75MW. This testing was made from July through August 1979, concluding 152 times measurements, 4 times PRM adjustments and twice Intermediate Range Monitors (IRM) adjustments, and followings were confirmed. (1)The PRM indicators show good linearity with the reactor thermal power. (2)The IRM indicators have non-linearity above 10MW, and exceedingly increase with the reactor power ascension. (3)The PRM indicators change with the graphite shield temperature changing, and the changing rate is +0.26 %/
C. The maximum changing width of the graphite shield temperature is about 40C and the changing behavior follows the reactor power history with time-constant of 2 or 3 days. The mechanism of this PRM indicators changing is the hardening of the transmitted neutron flux spectrum by arising of the graphite shield temperature. (4)The PRM indicators changes according to the reactor inlet sodium temperature, and the changing rate is +8.2
10
%/C. (5)The total error in the reactor thermal power measurement, which is including the systematic error and the random error, is 4.1MW (=5.5%) at 75MW rated power, while the random error is 0.40 MW (=0.53%). The error of the PRM indicators measurement is 0.42 %.
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PNC TN941 80-153, 46 Pages, 1980/09
This paper discribes the system design of the JOYO reactor containment atmosphere conditioning system and also includes the design which was revised because of the test results not satisfying the design requirement.
; Endo, Masayuki*; ; ; ; Sekiguchi, Yoshiyuki*
PNC TN941 79-179, 198 Pages, 1979/10
This report describes the results of the thermal power calibration test (PT-11) that was planned and performed as part of the power-up testing of the Experimental Fast Reactor "JOYO". The purpose of this test is to calibrate the Power Range Monitors (PRM) and Intermediate Range Monitors (IRM) by measuring the reactor thermal power at several levels from low power through the rated power of 50 MWt. The reactor thermal power was determined by measuring the inlet and outlet temperature and the flow rate of the primary main coolant. After this procedure, PRM and IRM were adjusted to coincide with the reactor thermal power by regulating the electronic amplifiers. Thes testing was made from April through August 1978, and followings were confirmed. (1)The PRM indicators show good linearity with the reactor thermal power. (2)The PRM indicators overlap with the IRM indicators over more than three decades. (3)The PRM indicators changes depending on the operating histry of the reactor. The PRM indicators is lower than the reactor thermal power immediately after start-up. Then that increases gradually and comes to stable in about one week after start-up. The maximum drifting value is about 6 per cent.
