Uncertainty Evaluation of Anticipated Transient without Scram Plant Response in the Monju Reactor Considering Reactivity Coefficients within the Design Range

• 作者: Masutake Sotsu Taira Hazama
• 来源期刊: 能源与动力工程:英文版
• 学科: 化学
• 关键词: FBR Monju ATWS uncertainty
• 年，卷（期）: 2019,（11）
• 页码范围: 393-403
• 页数: 11页
• 分类号: O64

摘要

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℃.