基于TRACE程序的华龙一号大破口失水事故现象分析

Analysis on Phenomena of HPR1000 Large Break Loss of Coolant Accident Based on TRACE Code

华龙一号(HPR1000)是我国自主研发的第3代核电技术,具有能动和非能动相结合、安注系统改进等设计特征。为分析上述华龙一号设计特征对大破口失水事故(LBLOCA)热工水力现象的影响,本研究基于TRACE程序,从核安全审评的角度对华龙一号大破口事故进程及其关键现象开展分析,对华龙一号、CPR1000和AP1000等机型在大破口失水事故进程和应对策略方面的差异进行了对比,并对事故不同阶段的主要热工水力现象和关键影响因素进行了分析和说明。结果表明,华龙一号核电厂LBLOCA中,事故进程的主要影响因素为破口喷放流量和安注箱背压,其事故序列与已有压水堆核电厂基本一致,基于计算结果识别的关键现象可为审评相关的现象识别与排序、模化分析、安全审评等提供技术支持和参考。

HPR1000 is the GenerationⅢpressurized water reactor(PWR)technology independently developed by China.In order to deal with the steam generator tube rupture(SGTR)accident and prevent pressurizer overflow,HPR1000 adopts some new features such as the combination of active and passive safety system,the reduction of pressure setting value of safety injection system,rapid cooling at the secondary side of steam generator.In order to analyze the impact of new design characteristics of HPR1000 on sequence and thermal-hydraulic phenomena in large break loss of coolant accident(LBLOCA),the numerical simulation of LBLOCA for HPR1000 was carried out using TRACE,which had been approved by United States Nuclear Regulatory Commission(NRC)as a best estimate system analysis code.The most challenging accident condition,namely combination of the most dangerous break location and the most dangerous size,were selected from the perspective of nuclear safety review.The LBLOCA sequence of HPR1000 was obtained and analyzed.The critical moments in the simulated LBLOCA process were compared with that of other typical commercial PWR such as CPR1000 and AP1000 in the sequence of accident and response strategies.According to the typical characteristics of thermal-hydraulic phenomena,the accident process was then divided into four stages,namely blowdown phase,refilling phase,reflooding phase and long term cooling phase.The main thermal-hydraulic phenomena in different accident stages except the long term cooling phase were identified and evaluated.The integral phenomena involved in the HPR1000 LBLOCA were depressurization of reactor coolant system(RCS),the coolant flow from core and intact RCS loop to broken loop,the safety injection flow and by-pass flow of accumulator(ACC).While the local phenomena were mainly blowdown flow at break,countercurrent flow limitations(CCFL)in downcomer and other channels with complex geometry,heat transfer in the core,two-phase flow and steam entrainment in the core,etc.The dominant factors during the accident process were pressure difference between the break and RCS,the pressure difference between ACC and RCS,core decay heat power and heat stored on the thick wall of RCS components.The results show that the main factors influencing the accident process are the mass flow rate of the break and the pressure setting value of accumulator in the LBLOCA of HPR1000.The accident sequence and phenomena are basically consistent with the existing commercial PWR nuclear power plants.The key phenomena identified based on the calculation results can provide technical support and reference for the phenomenon identification and ranking,scaling analysis,and nuclear safety review.