Theoretical analyses and design, construction and testing of a flow loop for the study of generalised forced and natural convection boiling heat transfer phenomena on typical light-water nuclear reactor fuel pin configurations

Abstract

In a worldwide pursuit for more Accident Tolerant nuclear Fuel (ATF), the quest to obtain and certify alternative nuclear fuel cladding tubes for light-water nuclear power reactors is still a key challenge. One of the facets in this program to develop more ATF is the heat transfer evaluation between the various proposed clad tubes manufactured from suitable replacement materials and the current problematic zirconium-alloy based clad tubes used in nuclear power reactors. For the heat transfer analysis, the accurate measurement of the temperature on the heat transfer surface of heated tubes to be tested was one of the important objectives for the effective analysis of the heat transfer characteristics to the water coolant. After extensive investigations, a suitable technique was developed and validated against recognised forced-convection heat transfer correlations. The results showed that this technique was well suited for external forced convection heat transfer studies from heated surfaces exposed to forced convection water coolants.