CFD-simulation of thermal hydraulic benchmark V1000CT–2 using ANSYS CFX

The OECD/NEA and AER coupled code benchmarks for light water reactors markedly contribute to the testing of best estimate coupled codes in reactivity transients. At the same time, these benchmarks and flow mixing studies indicate that further improvement of the mixing computation tools in the integrated codes is necessary.
Plant measured data from VVER-1000 coolant mixing experiments were used to test and validate CFD codes. The task is to compare the various calculations with measured data, using specified boundary conditions and core power distributions.
The experiments, which are provided for CFD validation, include single loop cooling down or heating-up by disturbing the heat transfer in the steam generator (SG) through the steam valves at low reactor power in the range of 5-14% and with all main coolant pumps (MCP) in operation. They were conducted during the plant commissioning phase at Kozloduy-6, Bulgaria and Kalinin-1, 2, Russia.
CFD calculations have been performed for the thermal hydraulic benchmark V1000CT-2 using ANSYS CFX. The numerical grid model was generated with the grid generator ICEM-CFD and contains 4.7 Mio. tetrahedral elements. The Best Practice Guidelines (BPG) in using CFD in Nuclear Reactor Safety Applications have been used. Different advanced turbulence models were used in the numerical simulation. The best agreement with the Kozloduy heating-up experiment at the core inlet shows the Detached Eddy Simulation (DES). Strong fluctuations occur in the downcomer of the RPV. The results show a clear sector formation of the affected loop at the downcomer, lower plenum and core inlet. The maximum local values of the relative temperature rise in the experiment amount 97.7% and in the calculation 97.3%. Uncertainties are still the estimation and interpolation of experimental values at the core outlet to the core inlet.