This paper describes the methods and results of an uncertainty evaluation of a significant plant response analysis of reactor trip failure events,specifically anticipated transients without scram in the Japanese proto...This paper describes the methods and results of an uncertainty evaluation of a significant plant response analysis of reactor trip failure events,specifically anticipated transients without scram in the Japanese prototype fast breeder reactor Monju.Unprotected loss of heat sink(ULOHS)has a relatively large contribution to the core damage frequency due to reactor trip failure.The uncertainty in the allowable time to core damage in this event has so far been estimated by considering the range of reactivity coefficients.There are some cases where it is considered that core damage will be avoided.Specifically,if the primary heat transport system(PHTS)pump inlet sodium temperature stays below 650℃for 1 h,the avoidance of core damage due to a ULOHS event is assumed.This is the temperature at which the probability of cavitation in the static pressure bearing begins to increase.In this study,a success scenario was investigated in two aspects:identification of influential input parameters and estimation of the probability of success.In the parameter identification,input parameters that satisfy the pump inlet temperature being below 650°C are clarified by treating the reactivity coefficients and reactor kinetics parameters as variables that can be taken to be within the design range.In the probability estimation,the results are fitted to a lognormal distribution function,from which the output variable was found to fall between 640 and 679℃with a probability of 90%,the probability of the temperature being 650℃or lower was 0.23,and the average and mode value was 659℃.展开更多
In order to evaluate the effects of the FHs (flow holes) on the inner barrel, which were installed in the upper plenum of the Monju reactor vessel, a high resolution meshes around the FHs was constructed. Using this...In order to evaluate the effects of the FHs (flow holes) on the inner barrel, which were installed in the upper plenum of the Monju reactor vessel, a high resolution meshes around the FHs was constructed. Using this model, it was mainly clear that in the 40% rated operational conditions, the shape of the FHs on the inner barrel did not change largely to the upper plenum thermal-hydraulics. The effect of the FHs on the honeycomb structure in the upper structure was also investigated in these calculations. The results indicated that the height of thermal stratification interface became lower than that evaluated from the test data.展开更多
The prototype fast breeder reactor "MONJU" has an EVSS (ex-vessel fuel storage system) which consists mainly of an EVST (ex-vessel fuel storage tank) and an EVST sodium cooling system. EVST sodium cooling system...The prototype fast breeder reactor "MONJU" has an EVSS (ex-vessel fuel storage system) which consists mainly of an EVST (ex-vessel fuel storage tank) and an EVST sodium cooling system. EVST sodium cooling system consists of three independent loops. During the normal operation, the primary sodium in the EVST is circulated by natural convection and the secondary circulation in the EVST sodium cooling system is powered by electromagnetic pumps. When an SBO (station blackout) occurs, all the pumps and blowers are tripped. Therefore, it was necessary to evaluate the cooling ability by the natural circulation of sodium in the EVST sodium cooling system and air through the air cooler during the SBO. In this study, an analysis and evaluation of the plant dynamics for the spent fuel and the EVSS structural integrity during an SBO were performed. When the number of cooling loops was not changed and natural circulation occurred in only two loops, the sodium temperature in the EVST increased to approximately 450 ~C. However, the structural integrity of the EVSS was maintained. The analytical results, therefore, help clarify the number of necessary cooling loops for efficient decay heat removal and sodium temperature behavior in an SBO.展开更多
In order to evaluate the operational reliability of Japanese FBR (fast breeder reactor) MONJU, frequencies of important intermediate events and equipment failures resulting during reactor automatic trip are predicte...In order to evaluate the operational reliability of Japanese FBR (fast breeder reactor) MONJU, frequencies of important intermediate events and equipment failures resulting during reactor automatic trip are predicted using FTA (fault tree analysis) technique for the plant system model. The targeted devices are the following: PHTS (primary heat transport system), SHTS (secondary heat transport system), WS (water and steam system), PPS (plant protection system) and PCS (plant control system). In this paper, the frequency of automatic reactor trips was estimated by extracting and analyzing the important intermediate events and equipment failures covering all the derived fault trees of these systems. The analyses predicted 1.2/RY (reactor year) the value of unplanned shut down frequency by the internal factor of the system. The largest contributed event was function failure of SHTS accounting for 42.6% of total events followed by PHTS with 40.1%. The contribution factor of WS was only 4.4%.展开更多
Thermal expansion behavior was investigated in detail for evaluation of the core support plate expansion reactivity in the ULOHS(Unprotected Loss of Heat Sink)reactor trip failure event.The core support plate expansio...Thermal expansion behavior was investigated in detail for evaluation of the core support plate expansion reactivity in the ULOHS(Unprotected Loss of Heat Sink)reactor trip failure event.The core support plate expansion reactivity plays an important role in the safety evaluation of the ULOHS event.In this paper,a possibility of mechanical restraint was investigated in thermal expansion of the core structure for the prototype FBR(Fast Breeder Reactor)Monju.The reactor core expansion was simulated in a three-dimensional FEA(Finite Element Analysis)model of the RV(Reactor Vessel)considering detailed temperature distribution of the sodium coolant based on the thermal-hydraulic analysis result of the whole core model.It was found that the thermal expansion of the core was not restrained in the ULOHS event,although part of the core structure is mechanically restrained.展开更多
文摘This paper describes the methods and results of an uncertainty evaluation of a significant plant response analysis of reactor trip failure events,specifically anticipated transients without scram in the Japanese prototype fast breeder reactor Monju.Unprotected loss of heat sink(ULOHS)has a relatively large contribution to the core damage frequency due to reactor trip failure.The uncertainty in the allowable time to core damage in this event has so far been estimated by considering the range of reactivity coefficients.There are some cases where it is considered that core damage will be avoided.Specifically,if the primary heat transport system(PHTS)pump inlet sodium temperature stays below 650℃for 1 h,the avoidance of core damage due to a ULOHS event is assumed.This is the temperature at which the probability of cavitation in the static pressure bearing begins to increase.In this study,a success scenario was investigated in two aspects:identification of influential input parameters and estimation of the probability of success.In the parameter identification,input parameters that satisfy the pump inlet temperature being below 650°C are clarified by treating the reactivity coefficients and reactor kinetics parameters as variables that can be taken to be within the design range.In the probability estimation,the results are fitted to a lognormal distribution function,from which the output variable was found to fall between 640 and 679℃with a probability of 90%,the probability of the temperature being 650℃or lower was 0.23,and the average and mode value was 659℃.
文摘In order to evaluate the effects of the FHs (flow holes) on the inner barrel, which were installed in the upper plenum of the Monju reactor vessel, a high resolution meshes around the FHs was constructed. Using this model, it was mainly clear that in the 40% rated operational conditions, the shape of the FHs on the inner barrel did not change largely to the upper plenum thermal-hydraulics. The effect of the FHs on the honeycomb structure in the upper structure was also investigated in these calculations. The results indicated that the height of thermal stratification interface became lower than that evaluated from the test data.
文摘The prototype fast breeder reactor "MONJU" has an EVSS (ex-vessel fuel storage system) which consists mainly of an EVST (ex-vessel fuel storage tank) and an EVST sodium cooling system. EVST sodium cooling system consists of three independent loops. During the normal operation, the primary sodium in the EVST is circulated by natural convection and the secondary circulation in the EVST sodium cooling system is powered by electromagnetic pumps. When an SBO (station blackout) occurs, all the pumps and blowers are tripped. Therefore, it was necessary to evaluate the cooling ability by the natural circulation of sodium in the EVST sodium cooling system and air through the air cooler during the SBO. In this study, an analysis and evaluation of the plant dynamics for the spent fuel and the EVSS structural integrity during an SBO were performed. When the number of cooling loops was not changed and natural circulation occurred in only two loops, the sodium temperature in the EVST increased to approximately 450 ~C. However, the structural integrity of the EVSS was maintained. The analytical results, therefore, help clarify the number of necessary cooling loops for efficient decay heat removal and sodium temperature behavior in an SBO.
文摘In order to evaluate the operational reliability of Japanese FBR (fast breeder reactor) MONJU, frequencies of important intermediate events and equipment failures resulting during reactor automatic trip are predicted using FTA (fault tree analysis) technique for the plant system model. The targeted devices are the following: PHTS (primary heat transport system), SHTS (secondary heat transport system), WS (water and steam system), PPS (plant protection system) and PCS (plant control system). In this paper, the frequency of automatic reactor trips was estimated by extracting and analyzing the important intermediate events and equipment failures covering all the derived fault trees of these systems. The analyses predicted 1.2/RY (reactor year) the value of unplanned shut down frequency by the internal factor of the system. The largest contributed event was function failure of SHTS accounting for 42.6% of total events followed by PHTS with 40.1%. The contribution factor of WS was only 4.4%.
基金The authors would like to recognize the contribution of Hiroki Yada for the thermal expansion analysis,and also Masaki Minami and Kousuke Araki of NESI for the thermal-hydraulic analysis in this paper.
文摘Thermal expansion behavior was investigated in detail for evaluation of the core support plate expansion reactivity in the ULOHS(Unprotected Loss of Heat Sink)reactor trip failure event.The core support plate expansion reactivity plays an important role in the safety evaluation of the ULOHS event.In this paper,a possibility of mechanical restraint was investigated in thermal expansion of the core structure for the prototype FBR(Fast Breeder Reactor)Monju.The reactor core expansion was simulated in a three-dimensional FEA(Finite Element Analysis)model of the RV(Reactor Vessel)considering detailed temperature distribution of the sodium coolant based on the thermal-hydraulic analysis result of the whole core model.It was found that the thermal expansion of the core was not restrained in the ULOHS event,although part of the core structure is mechanically restrained.