Study on Typical Design Basis Conditions of HPR1000 With Nuclear Safety Analysis Code ATHLET

Xi Huang,Weixin Zong,Chubin Lin,Yuan Yin,Zhikang Lin,Zhihao Ren,Ting Wang

doi:10.3389/fenrg.2020.00127

Abstract

The third-generation nuclear power plant Hua-long Pressurized Reactor (HPR1000) is developed based on the experience of Chinese commercial Nuclear Power Plant (NPP) designs, construction, operation and maintenance. It improves the concept of defense in depth and strengthens severe accident prevention and mitigation strategies. The HPR1000 has implemented a number of active and passive innovative safety systems and accident management procedures for design basis conditions, e.g., the employment of Medium Pressure Rapid Cooldown (MCD) and Atmospheric Steam Dump System (ASDS) for the activation of Middle Head Safety Injection (MHSI), the application of Secondary Passive Residual Heat Removal System (SPRHR) for the residual heat removal. In the article, calculations are carried out for HPR1000 nuclear power plant with nuclear system safety analysis code ATHLET (Analysis of Thermal-Hydraulics of Leaks and Transient) 3.1 [1]. By means of conservative deterministic safety analysis approach, transient analyses concerning selected typical design basis conditions, i.e., Large Break Loss-Of-Coolant Accident (LB-LOCA), Small Break Loss-Of-Coolant Accident (SB-LOCA), Steam Generator Tube Rupture accident (SGTR) and Feed water Line Break (FLB) are performed. The ATHLET results are also compared with the results performed by CGN (China General Nuclear) developed code LOCUST with similar assumptions. The comparisons indicate that, although some discrepancies are detected, the trends of system responses predicted by the two codes are generally in agreement with each other for different accident scenarios. The results also demonstrate that the acceptance criteria for each accident can be met with significant safety margin. Thus, the effectiveness of safety system configuration and accident management procedures is guaranteed.

Highlights

The Hua-long Pressurized Reactor (HPR1000) technology is developed based on experiences of Chinese commercial Nuclear Power Plant (NPP) designs, construction, operations, and maintenance (General Nuclear System Ltd., 2018a)
By considering similar initial conditions and assumptions for the simulation, the results performed with ATHLET are generally in good agreement with those predicted by LOCUST
For LOCUST, the water level experienced significant oscillation, but the pressure drop is not obvious. It can be inferred from these simulation results that for the Small Break Loss-Of-Coolant Accident (SB-LOCA) with above described penalizing assumptions, the configuration and operation strategy of the safety system ensures that the plant can be brought plant into a safe status, that the safety injection can be introduced into the primary system in time and it provides sufficient flow rate to guarantee the core to be covered

Summary

INTRODUCTION

The Hua-long Pressurized Reactor (HPR1000) technology is developed based on experiences of Chinese commercial Nuclear Power Plant (NPP) designs, construction, operations, and maintenance (General Nuclear System Ltd., 2018a). The study aims to investigate four typical DBC-3/4 conditions, i.e., LB-LOCA (DBC-4), SB-LOCA (DBC-3), SGTR (one tube) (DBC-3), and FLB (DBC-3) scenarios, which may result in relatively serious consequences and involve complicated thermal hydraulic phenomena e.g., the two phase critical break flow and two phase phenomena in reactor core These selected events can activate various safety systems, e.g., the Medium Pressure Rapid Cooldown (MCD) and may require further actions by the operators, and are of great concern for nuclear safety analysis and pertinent simulation tools. For LOCUST, the water level experienced significant oscillation, but the pressure drop is not obvious It can be inferred from these simulation results that for the SB-LOCA with above described penalizing assumptions, the configuration and operation strategy of the safety system ensures that the plant can be brought plant into a safe status, that the safety injection can be introduced into the primary system in time and it provides sufficient flow rate to guarantee the core to be covered. Discrepancies are detected between the results of ATHLET and LOCUST, the trends of system responses predicted by the two codes are generally in correspondence with each other

CONCLUDING REMARKS

DATA AVAILABILITY STATEMENT