AP1000主给水管道断裂事故中PRHR系统冷却能力分析

使用机理性分析程序建立包括主冷却剂系统、专设安全设施及相关二回路管道的AP1000核电厂模型，对AP1000核电厂主给水管道断裂事故进程进行计算分析。着重分析了非能动余热排出（PRHR）系统在主给水管道断裂事故工况中的瞬态响应、热工水力行为及其冷却能力，并针对PRHR系统流道阻力特性的不确定性对冷却能力的影响进行分析。分析结果表明，在主给水管道断裂事故中，PRHR系统的热移出功率最终能够与堆芯的衰变功率相匹配，有能力带走衰变热，保证一回路系统最终处于安全停堆状态，不发生堆芯损伤，当PRHR系统阻力系数增加时，PRHR系统的流量和换热功率会降低，对PRHR系统冷却能力造成影响。

Analysis on PRHR System Cooling Capacity During Feedwater System Pipe Break Accident for AP1000

AP1000 plant model was established, which included the reactor coolant system (RCS), engineering safety features (ESF), and related second side pipes system. Feedwater system pipe break accident of AP1000 plant was selected to analyze the accident progression. The transient response, thermal-hydraulic phenomena and the cooling capacity of the passive residual heat removal (PRHR) system was focused on. And the impact of pipe resistance uncertainty on PRHR system cooling ability was analyzed. The results show that the removal heat power of PRHR system heat exchanger can match with the core decay power at the late stage of the accident, and meets the requirements of sustainable long-term core cooling in case of normal residual heat removal path failure. When PRHR system resistance coefficient increases, PRHR system flow rate and heat transfer power decreases, and PRHR system cooling ability will be affected.