Analysis of steam generator tube rupture accident for OPR 1000 nuclear power plant

Abstract

Evaluation of radionuclide releases to the environment from operating nuclear power plants has increased in importance in the Republic of Korea following the 2015 amendment to its Nuclear Safety Act that includes not only design basis accidents but also severe accident management. Containment bypass accidents are among the crucial scenarios that should be considered under the new circumstances due to their high consequences, even though the probability of such accidents is low. In this study, analysis of a thermally induced steam generator tube rupture (TI-SGTR) accident caused by a severe accident has been conducted with MELCOR. To increase the reliability of the analysis results, qualification processes were performed in the process of producing the input. In addition, computational fluid dynamics analysis was performed to simulate the natural circulation phenomena in the reactor coolant system (RCS) during the accident and major flow parameters were obtained for the target plant. Thermal-hydraulic and fission product behavior was found during the accident for the base case scenario, and uncertainty band for CsOH release fraction to the environment were presented through the uncertainty analysis on the natural circulation flow parameters. Sensitivity calculations were also performed on uncertain parameters such as the number of ruptured tubes and the depressurization time of the secondary system, and the CsOH release fraction to the environment without passing through the containment were evaluated.