Thermal-Hydraulic Behavior and Safety Characteristics of the NHR200-II Reactor During Small-Break Loss-of-Coolant Accidents

The development of advanced integrated small reactors is a significant trend in modern nuclear engineering. Designed by the Institute of Nuclear and New Energy Technology at Tsinghua University, NHR200-II is a nuclear heating reactor that incorporates a range of advanced passive and inherent safety features. Different from general pressurized water reactors, the safety injection system is excluded from the NHR200-II design to enhance system safety and simplicity. Therefore, the potential occurrence of a loss-of-coolant accident (LOCA) will be a key concern in the NHR200-II reactor design. The main objective of this study is to develop a Reactor Excursion and Leak Analysis Program 5 (RELAP5) model suitable for small-break loss-of-coolant accident (SB-LOCA) analysis and to explore the dynamic characteristics of system thermal-hydraulic parameters during the accident, as well as to assess the safety of NHR200-II. In this paper, a detailed thermal-hydraulic analysis of two typical SB-LOCA scenarios is presented: (1) pipe rupture of the hydraulic control rod driving system with isolation failure and (2) rupture of a small-diameter pipe on the head of the reactor pressure vessel (RPV). Several key safety characteristics, including total coolant loss from the RPV, liquid level remaining in the RPV, and maximum temperature of fuel rods, are examined to assess the safety performance of NHR200-II. The results indicate that in both accident scenarios, the remaining coolant in the primary loop is adequate to cover the reactor core and effectively remove decay heat via the passive residual heat removal system, which prevents the reactor core from overheating damage and ensures the reactor remains in a safely controllable state. Thus, the safety characteristics of the NHR200-II reactor in LOCAs are demonstrated.