Effect of Rod Internal Pressure on Simulation of Halden Test IFA-650.9 with FE-based Fuel Analysis Code MERCURY

Analysis for the behavior of nuclear fuel in loss of coolant accidents in a nuclear reactor is being emphasized from the viewpoint of the characteristics of the nuclear fuel materials and the ability to maintain the coolable geometry. In particular, a nuclear fuel performance code based on the finite element method is being developed for the realistic nuclear fuel analysis. As a representative code, the finite element methodbased BISON code [1] for the evaluation of nuclear fuel performance under steady-state and accident conditions has been developed in the CASL project of the US and the version is being continuously updated. In KAERI, a multi-dimensional code development project for nuclear fuel analysis of accident conditions is being carried out, and a code named MERCURY [2] is being developed. In the process of developing the analysis code, it is necessary to validation for the various experimental data with the verification process. In this paper, the analysis results are presented in terms of the rod internal pressure of the MERCURY code for IFA-650.9, which is one of the IFA-650 series [3], which is the LOCA simulation test of the Halden Research Reactor for inpile test data.