Abstract: 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 ℃.