The Level-2 probabilistic safety assessment(PSA) of pressurized water reactors studies the possibility of creep rupture for major reactor coolant system components during the course of high pressure severe accident se...The Level-2 probabilistic safety assessment(PSA) of pressurized water reactors studies the possibility of creep rupture for major reactor coolant system components during the course of high pressure severe accident sequences. The present paper covers this technical issue and tries to quantify its associated phenomenological uncertainties for the development of Level-2 PSA. A framework is proposed for the formal quantification of uncertainties in the Level-2 PSA model of a PWR type nuclear power plant using an integrated deterministic and PSA approach. This is demonstrated for estimation of creep rupture failure probability in station blackout severe accident of a 2-loop PWR, which is the representative case for high pressure sequences. MELCOR 1.8.6 code is employed here as the deterministic tool for the assessment of physical phenomena in the course of accident. In addition, a MATLAB code is developed for quantification of the probabilistic part by treating the uncertainties through separation of aleatory and epistemic sources of uncertainty.The probability for steam generator tube creep rupture is estimated at 0.17.展开更多
文摘在核反应堆发生严重事故时,自然对流、传热和湍流现象对安全壳内的气体流动和分布起着重要作用。为研究事故发生时安全壳内气体流动和换热现象,并分析自然对流下的温度分布。采用计算流体力学(Computational Fluid Dynamics,CFD)方法开展数值模拟研究,首先利用差异加热空腔实验数据对CFD方法分析封闭空间内自然对流换热现象的适用性进行验证,然后建立了与THAI(Thermal-hydraulics,Hydrogen,Aerosols and Iodine)-TH21实验装置相对应的几何模型,分析了实验过程中安全壳内压力变化以及达到稳定状态后装置内的流动换热特性。结果表明:本文采用的计算方法对分析安全壳内气体流动换热具有较好的适用性,模拟得到的准静态压力高于实验压力1.61%,冷壁面附近温度分布与实验数据吻合良好,热壁面附近温度分布误差在2%以内。研究有助于相关数值方法与模型的验证与开发,并可为反应堆事故发生时安全壳内气体流动及分布提供一定的参考。
文摘The Level-2 probabilistic safety assessment(PSA) of pressurized water reactors studies the possibility of creep rupture for major reactor coolant system components during the course of high pressure severe accident sequences. The present paper covers this technical issue and tries to quantify its associated phenomenological uncertainties for the development of Level-2 PSA. A framework is proposed for the formal quantification of uncertainties in the Level-2 PSA model of a PWR type nuclear power plant using an integrated deterministic and PSA approach. This is demonstrated for estimation of creep rupture failure probability in station blackout severe accident of a 2-loop PWR, which is the representative case for high pressure sequences. MELCOR 1.8.6 code is employed here as the deterministic tool for the assessment of physical phenomena in the course of accident. In addition, a MATLAB code is developed for quantification of the probabilistic part by treating the uncertainties through separation of aleatory and epistemic sources of uncertainty.The probability for steam generator tube creep rupture is estimated at 0.17.
