STUDY ON THE BREAK ACCIDENTS OF THE HTR-PM PRIMARY LOOP

摘要

In thermal hydraulics designing and safety analysis of the High Temperature gas-cooled Reactor-Pebble Bed (HTR-PM), the THERMIX code was used to study the behavior of helium in the primary coolant system. Once the helium leaks out of the primary loop through a break on the pressure boundary or an inadvertent open relief valve, it is difficult to simulate the conditions of the room where the release occurred with THERMIX. In this paper, the latest version of RELAP5/MOD4 was used to simulate the behavior of the helium released to the containment rooms. A RELAP5/MOD4 input deck of the HTR-PM, consisting of the core, the primary coolant system, the secondary loop and the containment, was developed and evaluated in this paper. Based on the model, this paper simulated the accidents consequences of large breaks or small breaks near the inlet or the outlet of the helium circulator located inside the steam generator pressure vessel. The calculating results illustrate that the temperature of the helium flowing into the reactor building through the break was no more than 280°C even after an un-isolating large break. The analysis shows that the systems function to scram the reactor and to monitor the core temperature and pressure after accidents would not be affected by breaks.