Thermalhydraulic behavior of electrically heated rod during a critical heat flux transient

Abstract

In nuclear reactors, the occurrence of critical heat flux leads to fuel rod overheating with clad fusion and radioactive products leakage. To predict the effects of such phenomenon, experiments are performed using electrically heated rods to simulate operational and accidental conditions of nuclear fuel rods. In the present work, it is performed a theoretical analysis of the drying and rewetting front propagation during a critical heat flux experiment, starting with the application of an electrical power step from steady state condition. After the occurrence of critical heat flux, the drying front propagation is predicted. After a few seconds, a power cut is considered and the rewetting front behavior is analytically observed. Studies performed with various values of coolant mass flow rate show that this variable has more influence on the drying front velocity than on the rewetting one

Investigation of Mixing Flow Process Using PIV and PLIF Techniques

“…The power generation of a nuclear reactor is limited by the coolant capability of removing the heat generated inside the fuel rods. In order to avoid the occurrence critical heat flux (CHF) the deviation between the average outlet temperature of fuel assembly and temperature measured with in core thermocouples must be controlled [1]. In some cases the measurements can only provide limited information at specific locations where the tools could be placed.…”