Abstract
For analysis of Beyond Design Basis Event, PSA methods would be initiated to determine event sequences’ frequency of core damage. However, its estimation utilizing large numbers of conditional assumptions and criteria don’t have sufficient consideration of parameters uncertainty. RISMC method which addresses aleatory uncertainty would generate different result compared with traditional way. In light of both of sequences’ damage and frequency, this research uses Small Break Loss of Coolant Accident event sequence S(2)S20, 3rd of top ten internal initiating event sequences with highest core damage frequency over 1.0E-7/year ranked in Probabilistic Risk Assessment Maanshan Nuclear Power Station Final Report, as example (which 1st and 2nd are Loss of Safety 4.16kV Bus A-PB-S02 and Break of Steam Generator Tube respectively) to evaluate degree of difference between two methods. Although it already has a comprehensive estimation of Human Reliability Conditional Exceedance Probability in Maanshan Nuclear Power Plant Human Reliability Report, the actual operation time would be extremely affected if taking aleatory uncertainty as consideration. This research is expected to use RELAP5-3D to simulate SBLOCA event sequence with both plant operation factor and human factor parameters uncertainty. Then utilize statistical methods to approach probability distribution function, calculating operation success rate in step of Core Cooling Recovery. The result would be compared with past data in safety report.