Application of Thermal Hydraulic and Severe Accident Code SOCRAT/V3 to Bottom Water Reflood Experiment QUENCH LOCA 1

Abstract

This study aims to (1)آ use the thermal hydraulic and severe fuel damage (SFD) bestestimate computer modeling code SOCRAT/V3 for posttest calculation of QUENCHLOCA1 experiment and (2)آ estimate the SOCRAT code quality of modeling. The new QUENCHLOCA bundle tests with different cladding materials will simulate a representative scenario for a lossofcoolantaccident (LOCA) nuclear power plant (NPP) accident sequence in which the overheated (up to 1050آ°C) reactor core would be reflooded from the bottom by the emergency core cooling system (ECCS). The test QUENCHLOCA1 was successfully performed at the KIT, Karlsruhe, Germany, on February 2, 2012, and was the first test for this series after the commissioning test QUENCHLOCA0 conducted earlier. The SOCRAT/V3calculated results describing thermal hydraulic, hydrogen generation, and thermomechanical behavior including rods ballooning and burst are in reasonable agreement with the experimental data. The results demonstrate the SOCRAT code’s ability for realistic calculation of complicated LOCA scenarios.