407.1 Post-Test Calculation of the QUENCH-17 Bundle Experiment with Debris Formation and Bottom Water Reflood Using Thermal Hydraulic and Severe Fuel Damage Code SOCRAT/V3 Alexander Vasiliev Nuclear Safety Institute (IBRAE) B.Tulskaya 52 115191 Moscow, Russia vasil@ibrae.ac.ru Juri Stuckert Karlsruhe Institute of Technology (KIT) Kaiserstrasse 12 76131 Karlsruhe, Germany juri.stuckert@kit.edu ABSTRACT The thermal hydraulic and SFD (Severe Fuel Damage) best estimate computer modelling code SOCRAT/V3 was used for the calculation of the QUENCH-17 experiment which was the first test in the QUENCH tests series simulating debris behaviour. The QUENCH-17 test conditions simulated a representative scenario of nuclear power plant severe accident sequence with debris bed formation in which the overheated up to 1800 K core would be reflooded from the bottom by ECCS (Emergency Core Cooling System). The QUENCH-17 test included the following phases:  Heat-up phase (heat-up rate up to 0.25 K/s);  Oxidation phase (the cladding temperature T1800 K in hottest region, steam mass flow rate 2 g/s);  Bottom flood phase (characteristic cooling time  600 s, water mass flow rate 10 g/s). The test QUENCH-17 was successfully conducted at the KIT, Karlsruhe, Germany, on January 30-31, 2013. The objective of this test was to examine the formation of a debris bed inside the completely oxidised region of the bundle without melt formation and to investigate the coolability behaviour during the reflood. The QUENCH facility is designed for studies of the PWR fuel assemblies behaviour under conditions simulating design basis, beyond design basis and severe accidents. The test bundle for the QUENCH-17 test was intentionally changed in comparison to basic QUENCH tests with the emphasis to investigate debris behaviour phenomena. Only 12 periphery fuel rod simulators were heated. 9 unheated fuel rod simulators were located in the inner part of the test bundle. This is why the massive porous debris formation in the inner part of the bundle was not influenced by the presence of tungsten heaters. The SOCRAT/V3 computer modelling code was used for the calculation of basic thermal hydraulic, oxidation and thermal mechanical behaviour during all phases of the experiment. The calculated results are in a good agreement with experimental data which justifies the adequacy of modelling capabilities of the SOCRAT code system.