Modeling of Thermal Hydraulics Aspects of Combined Top and Bottom Water Reflood Experiment PARAMETER-SF2 Using SOCRAT 2.1 Code

Abstract

The PARAMETER-SF2 test conditions simulated a severe LOCA (Loss of Coolant Accident) nuclear power plant sequence in which the overheated up to 1700÷2300K core would be reflooded from the top and the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in April 3, 2007 and was the second of two experiments to be performed in the frame of ISTC 3194 Project. PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents. After the maximum cladding temperature of 1750K was reached in the bundle during PARAMETER-SF2 test, the top flooding (flow rate 40g/s) was begun and later approximately in 30 s the bottom flooding (flow rate 100g/s) was initiated. Two-phase (water and steam) flow determined the fuel assembly cooling conditions. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate numerical complex SOCRAT 2.1 was used for the calculation of PARAMETER-SF2 experiment. Thermal hydraulics in PARAMETER-SF2 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT 2.1 were compared with experimental data concerning different aspects of thermal hydraulics behavior including convective and radiative heat transfer in the bundle and the CCFL (counter-current flooding limitation) phenomenon during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF2 test.