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
Creators
- 1. Nuclear Safety Institute (IBRAE), Moscow (Russian Federation)
- 2. Karlsruher Institut of Technology (KIT), Karlsruhe (Germany)
Description
The thermal hydraulic and SFD (Severe Fuel Damage) best estimate computer modelling code SOCRAT/V3 was used for the calculation of 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 1800K 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 > 1,800 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. 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 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 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.
Additional details
Publishing Information
- Journal Title
- Atw. Internationale Zeitschrift fuer Kernenergie
- Journal Volume
- 59
- Journal Issue
- 11
- Journal Page Range
- p. 621-630
- ISSN
- 1431-5254
INIS
- Country of Publication
- Germany
- Country of Input or Organization
- Germany
- INIS RN
- 51085279
- Subject category
- S22: GENERAL STUDIES OF NUCLEAR REACTORS;
- Descriptors DEI
- COMPUTERIZED SIMULATION; CONTINUITY EQUATIONS; CONVECTION; DESIGN-BASIS ACCIDENTS; ECCS; EQUATIONS OF MOTION; FLOW RATE; FUEL ELEMENT CLUSTERS; GAS FLOW; LIQUID FLOW; NUCLEAR POWER PLANTS; OXIDATION; POROUS MATERIALS; PWR TYPE REACTORS; S CODES; STEAM; THERMAL HYDRAULICS; TIME DEPENDENCE; WATER
- Descriptors DEC
- ACCIDENTS; CHEMICAL REACTIONS; COMPUTER CODES; DIFFERENTIAL EQUATIONS; ENERGY TRANSFER; ENGINEERED SAFETY SYSTEMS; ENRICHED URANIUM REACTORS; EQUATIONS; FLUID FLOW; FLUID MECHANICS; FUEL ASSEMBLIES; HEAT TRANSFER; HYDRAULICS; HYDROGEN COMPOUNDS; MASS TRANSFER; MATERIALS; MECHANICS; NUCLEAR FACILITIES; OXYGEN COMPOUNDS; PARTIAL DIFFERENTIAL EQUATIONS; POWER PLANTS; POWER REACTORS; REACTOR PROTECTION SYSTEMS; REACTORS; SIMULATION; THERMAL POWER PLANTS; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Notes
- This record replaces 46000049