核电厂大破口事故建模与初始参数不确定性分析

Large-break loss of coolant accident modeling and the uncertainty analysis of initial parameters

以AP1000核电厂为原型,利用系统程序RELAP5建模模拟AP1000大破口失水事故,并与西屋公司大破口失水事故分析结果进行比较,另采用数学分析与灵敏度分析方法对电厂初始参数进行不确定性量化分析.比较结果显示:RELAP5和西屋公司的LBLOCA(large-break loss of coolant accident)计算结果有较好的一致性,而由数学分析和灵敏度分析处理电厂重要状态参数不确定性后,相对于保守的电厂参数包络LOCA(loss of coolant accident)分析,能额外提供30～50K的热工裕量.

In this paper, AP1000 nuclear power plant is simulated by system code RELAP5 and AP1000 LBLOCA （large-break loss of coolant accident） is analyzed. Associated analysis provides contrast to the LBLOCA of Westinghouse AP1000. Besides, quantification of the effect of plant status uncertainty on safety parameter is performed based on mathematic statistics and sensitivity analysis. The results show that after sensitivity analysis and quantification of the plant status uncertainty, a greater PCT （peak cladding temperature） margin can be provided by 30-50 K than traditional conservative bounding status LOCA （loss of coolant accident） analysis.