基于PSA的压水堆LBLOCA不确定性分析

为了结合确定论与概率论分析开展更加真实的核反应堆事故工况安全分析，提出了一种结合概率安全分析（PSA）和最佳估算加不确定性（BEPU）分析的方法，并以典型三环路压水堆冷管段双端断裂大破口失水事故（LBLOCA）的极限事故为对象，首先基于PSA开展了应急堆芯冷却系统的事故失效分析，而后结合BEPU分析评估了事件树中各事故序列的包壳峰值温度（PCT）分布及条件堆芯损坏概率（CCDP），最终确定了压水堆在该事故工况中的堆芯损坏频率（CDF）。分析结果表明，压水堆在冷管段双端断裂工况中应急堆芯冷却系统能够保证反应堆的安全，且一列低压安注系统足以排出堆芯余热及保证反应堆安全。 

PSA-Based Uncertainty Analysis of Pressurized Water Reactor LBLOCA

In order to combine deterministic and probabilistic analysis to carry out more realistic safety analysis of nuclear reactor accident conditions, a method combining probabilistic safety analysis (PSA) and best estimate plus uncertainty (BEPU) analysis is proposed, the extreme accident of the double-ended fracture large-break loss of water accident (LBLOCA) in a typical three-loop pressurized water reactor cold pipe section is taken as the object. First, the accident failure analysis of the emergency core cooling system is carried out based on PSA. Then, combined with BEPU analysis, the cladding peak temperature (PCT) distribution and conditional core damage probability (CCDP) of each accident sequence in the event tree are evaluated, and the core damage frequency of the PWR in this accident condition is finally determined. The analysis results show that the emergency core cooling system of PWR can ensure the safety of the reactor under the condition of double-ended fracture of cold pipe section, and a row of low-pressure safety injection systems are sufficient to remove the residual heat of the core and ensure the safety of the reactor.