Probabilistic Safety Assessment For SGTR Accident In AP1000 Nuclear Power Plant

Probabilistic Safety Assessment (PSA) is differ from traditional Deterministic Safety Analysis, PSA can provide comprehensive analysis for complicated systems, which has various kinds of development processes. If accident happens, then, the accident sequences caused by different initial events is acquired, so that accident frequency and consequences can be analyzed systematacially, more importantly, the potential cause of NPP accidents can be found.AP1000 is an advanced pressurized nuclear power plant with Generation III reactor, its passive design concept is well-received. In this paper, The Steam Generator Tube Rupture(SGTR) accident is analyzed based on the PSA report of self-independent standard design,the Chinese version of Sanmen nuclear power plant PSAR preliminary safety analysis report and some corresponding references and so on.Firstly, in this paper, the AP1000 structure has been introduced, accident progress of SGTR and response actions of relief systems have been analyzed.Secondly, in the paper, the event tree model has been built to deduce the accident progress, then the systems relate to this accident have been modeled by fault tree method.Also, the common cause failure models and groups have been built in detail, more than that,the human reliability about the fault trees have been disassembled to be several tasks with descriptions.Finally, the Risk Spectrum software has been used to finish the quantitative and qualitative analysis of the system fault trees, the consequences of failure unavailability and minimal cut sets of system fault tree have been acquired. The total core damage frequency caused by SGTR accident has been obtained through connecting fault trees and event trees,the importance measures, sensibility and uncertainty of core damage have also been analyzed respectively. It turned out that the core damage frequency (CDF) caused by SGTR accident is 3.95×10-9 per reactor year. The lower(5%) and upper(95%) limit of confidence interval are 6.22×10-11 per reactor year and 12.71×10-8 per reactor year. Sensitivity analysis shows that total failure of human actions have great impact to CDF, but reducing human error cannot bring good benefit for the mitigation of SGTR, furthermore, ADS and IRWST are the most important systems in the sensitivity case.