压水堆核电站大破口失水事故分析

压水堆核电站安全分析报告是核安全监管部门对其进行安全审查的重要文件，大破口失水事故是核电站运行的设计基准事故，是安全分析报告中的重要内容。本文使用RELAP5/MOD3.2进行压水堆冷管段大破口失水事故的计算，对比发现一回路冷管段发生双端断裂大破口时燃料元件包壳温度峰值（PCT）最高，且长时间维持在较高温度，此条件下反应堆最危险。计算结果表明，事故发生后，一回路压力迅速下降，堆芯冷却剂的流动性变差，导致堆芯裸露，燃料包壳温度又重新回升。通过安注系统和辅助给水系统等一系列动作，能保证燃料元件包壳温度不超过1204 ℃的限值。

Safety Analysis of Large Break Loss of Coolant Accident of PWR Nuclear Power Plant

The safety analysis report of the pressurized water reactor (PWR) nuclear power plant is an important document for the safety review of the nuclear safety supervision department. The large break loss of coolant accident is a design basis accident for the operation of nuclear power plants and an important part of the safety analysis report. In this paper, RELAP5/MOD3.2 was used to calculate the large break loss of coolant accident of the PWR cold-leg section. It is found that the peak of fuel element cladding temperature (PCT) is the highest when the double-end fracture occurs in the cold-leg section of the primary loop and maintains at a higher temperature for a long time, so the reactor is the most dangerous in this condition. The calculation results show that the pressure of the primary loop drops rapidly after the double-end fracture accident, and the fluidity of the core coolant deteriorates, resulting in the core exposed and the fuel cladding temperature rising again. Through a series of actions such as an injection system and an auxiliary water supply system, it is possible to ensure that the fuel element cladding temperature does not exceed the limit of 1204 ℃.