Thermal-hydraulics validation of CFD codes for Light Water Nuclear Reactors against benchmark experimental results

The cooling of a nuclear reactor depends of a suitable fluid flow pattern among its fuel elements aiming the removal of heat produced in the fuel. In case of light water reactors, an excess of heat drives the fluid to change from liquid to vapor, reducing significantly its cooling capacity and leading the reactor to a LOCA accident. A numerical CFD simulation code is a suitable tool that allows to study this kind of problem and explore the limits that should be avoided during reactors operation. The CFD simulations could be performed by dedicated codes or commercial codes. This last, had proven to be able to simulate with high accuracy and confidence a sort of equipment or systems, avoiding spending efforts and financial resources for development new codes that, essentially, do the same. In any case, the CFD codes must be validated against experimental results. To comply with this objective, a benchmark fuel element was purposed and experimentally essayed to provide experimental results for CFD codes calibration. The results of this essay are provided to the four types of subchannels for a 5x5 PWR fuel element, with results provided as density and void fraction. The present work aims to present the preliminary results obtained with CFD numerical simulations using the ANSYS-CFX® code for the central subchannel with active rods for stead state operation. The results demonstrated that the CFX® is adequate to simulate with high accuracy the flow in this subchannel. (author)