压水堆大破口失水事故重要现象识别及数值计算不确定性量化分析研究

Study on Important Phenomenon Identification and Numerical Simulation Uncertainty Analysis for PWR Large Break LOCA

大破口失水事故是压水堆核电厂最重要的设计基准事故,对该事故的准确模拟可为提升反应堆功率提供重要支撑。本文采用最佳估算程序RELAP5对压水堆失水事故试验(LOFT)的实验工况FP-LP-2进行了模拟计算,并应用德国反应堆安全研究所(GRS)不确定性分析方法对计算结果进行不确定性量化和敏感性分析;给出了关键输出参数95%置信度的不确定性包络带,并分析了计算结果的不确定性变化趋势及原因。分析结果表明,对包壳峰值温度影响较大的重要现象包括堆芯衰变热、完整环路破口临界流喷放系数和燃料棒的热导率。本文研究确认了GRS方法的有效性,为改进现有核电站安全分析方法具有积极作用。

Large break loss of coolant accident is one of the most important design basis accidents, the accurate calculation can provide great support to enhance the plant power. In this study, the best estimate program RELAP5 is used to simulate the condition FP-LP-2 during the experiments of the loss of fluid test(LOFT), and the Gesellschaft für Anlagen-und Reaktorsicherheit(GRS) uncertainty analysis method is applied to the uncertainty quantification and sensitivity analysis. The uncertainty envelope with 95% confidence of key output parameters is given, and the uncertainty trend of these output results and their reasons are analyzed. Sensitivity analysis shows that the factors with effects on the peak cladding temperature include core decay heat, the critical flow discharge coefficient of the intact loop break, and the thermal conductivity of the fuel rod. This study confirmed the effectiveness of the GRS method, and can provide support to improve the safety analysis method.