Structural Integrity Evaluation of a Reactor Cavity during a Steam Explosion for External Reactor Vessel Cooling

Sang-Hyun Park,Jong-Rae Cho,Kwang-Hyun Bang

doi:10.3390/en14123605

Abstract

Nuclear power is a major source of electricity in the international community. However, a significant problem with nuclear power is that, if a severe nuclear accident occurs, radiation may leak and cause great damage. As such, research on nuclear safety has become increasingly popular worldwide. In this paper, the structural integrity of a reactor cavity during a steam explosion—one kind of the aforementioned severe nuclear accidents—was evaluated. Steam explosions are primarily caused by fuel–coolant interactions (FCI), and result from issues in the cooling system that discharges the melt from the reactor core to the outside. A steam explosion can damage the nuclear power plant, and radiation leakage, the greatest concern, may occur. In the Chernobyl or Fukushima Daiichi accidents, significant radiation leakages resulted in damages extending beyond the country of origin. In this paper, a steam explosion was simulated using values given by the transient analysis code for explosive reactions (TRACER-II)—the only steam explosion code in Korea. The walls of the reactor cavity were modeled after the APR-1400 currently operating in Korea. The integrity of the concrete, rebars, and liner plate in the reactor cavity during a steam explosion was evaluated in terms of stress and ductile failure strain limits.

Highlights

According to the International Nuclear and Radiological Event Scale (INES) [1], a severe accident is defined as one that goes beyond a design-basis accident and causes core damage
The maximum principal stress over time of the concrete used in the walls of the reactor analysis, the effect of tension was examined based on the primary maximum principal stress
The structural integrity of the walls of a reactor cavity during a steam explosion was evaluated for IVR-ERVC

Summary

Introduction

According to the International Nuclear and Radiological Event Scale (INES) [1], a severe accident is defined as one that goes beyond a design-basis accident and causes core damage. If the decay heat in the core cannot be reduced owing to complications in the coolant system, a temperature rise may melt and damage the cladding, causing a severe accident. A typical example is the nuclear meltdown that occurred in 1979 at the Three Mile Island Unit 2 reactor, Pennsylvania. This accident was caused by a core meltdown due to a violation of the procedure guidelines, defects in the facility, and repeated operator mistakes [2]. Designs that reinforced human-machine connectivity have been implemented as follow-up measures

Objectives

Results

Conclusion