AP1000丧失正常给水事故PRHR冷却能力研究

 AP1000作为第三代革新型核电厂,广泛采用了非能动安全设计,来提高系统的安全性和经济性。其中,非能动余热排出系统(PRHR)用于应对正常余热排出路径失效的事故。本文采用机理性分析程序建立了包括主冷却剂系统(RCS)、专设安全设施(ESF)、以及简化的二回路系统的AP1000核电厂模型,对AP1000核电厂丧失正常给水事故进程进行了模拟计算。着重分析了非能动余热排出系统在丧失正常给水事故工况中的瞬态响应、热工水力行为及其冷却能力,并将PRHR与内置换料水箱(IRWST)的换热功率与堆芯衰变热功率进行了比较。研究表明,在丧失正常给水事故中,PRHR的热移出功率最终能够与堆芯的衰变功率相匹配,PRHR热交换器(PRHR HX)有能力带走衰变热,将反应堆主系统维持在安全停堆的状态。

The PRHR Cooling Capacity during the Fail of Normal Feedwater Flow Accident for AP1000

AP1000 is the third generation of innovation-typed nuclear power plant and widely adopts passive safety design to increase safety and efficiency. Particularly, a Passive Residual Heat Removal (PRHR) system is used to deal with the accident due to the failure of normal residual heat removal path. AP1000 plant model is established by using mechanism-based code. The plant model includes the Reactor Coolant System (RCS), Engineering Safety Features (ESFs), and part simplified second side system. And the fail of normal feedwater flow accident of AP1000 plant is selected to analyze the accident progression. The transient response, thermal-hydraulic phenomena, and the cooling capacity of the PRHR are focused on. The transfer heat power from PRHR to IRWST and reactor core decay power are compared with each other. The result shows that in the case of failure accident due to normal residual heat removal path, the removal heat power of PRHR Heat eXchanger (PRHR HX) is able to match with the reactor core decay power at the late stage of the accident, meets the requirements of sustainable long-term core cooling, and keeps the reactor system in the safe shutdown situation.