Simulation of high temperature fuel rod bundle behavior during the QUENCH-06 experiment using CINEMA code

Abstract

The CINEMA code has been developed to secure the ability to analyze severe accidents in nuclear power plants with a Korean domestic computer code. As part of the code validation, the current work simulated the QUENCH-06 experiment with the CINEMA. Calculation results from CINEMA were compared to the experimental results and numerical results of SPACE and SCDAP/RELAP5. CINEMA overestimated the bundle temperatures near the test section entrance, as observed in the SPACE and SCDAP/RELAP5 calculation due to an underestimation of the thermal conductivity of argon-filled ZrO2 fiber insulation. The hydrogen generation rate was overestimated by the CINEMA because of the high bundle temperature at the upper part, where CINEMA is unable to model extreme heat loss through the shroud without insulation. The CINEMA results demonstrated good agreement with the experimental data, validating that the CINEMA is capable of simulating high temperature fuel rod behavior before and during the reflood phase.