Thermal-hydraulics validation of CFD code for light water nuclear reactors against benchmark experimental results

Abstract

The cooling of a nuclear reactor depends on 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 its phase from liquid to vapor, significantly reducing its capacity to remove heat and leading the reactor to a Loss of Coolant Accident. Numerical simulations using a CFD code is a suitable tool to address this kind of problem and explore the conditions that should be avoided during the reactor operation. The commercial CFD codes had proven to be reliable to simulate with a high accuracy and confidence the thermal-hydraulics of a sort of equipment and systems, avoiding spending efforts and financial resources in the development of new codes that, essentially, perform the same tasks. Despite of it, the CFD codes must be validated, such as 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. This work presentes 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 ANSYS-CFX^® is adequate to simulate with high accuracy the flow in this subchannel.