Abstract
The transient behavior of nuclear fuel during loss of coolant accident (LOCA) must be evaluated not only for power reactors but also for research reactors. Kyoto University Research Reactor (KUR) is a light-water moderated tank-type reactor operated at the rated thermal power of 5MW and has been widely utilized for a lot of scientific researches. The safety of the plate-type fuels in the KUR has been studied by a reactor coolant system analysis code. However, for the accurate prediction of the natural convection phenomena of air during LOCA, the empirical correlation for natural convection should be selected properly and an experimental validation should be performed. In this study, the natural convection heat transfer characteristics were experimentally investigated using a simulated fuel assembly which has the same dimension with the KUR fuels. Furthermore, a simplified unsteady heat conduction simulation was performed by taking the axial heat density distribution into account, and the cooling feature of the fuel plate during LOCA was discussed by estimating the time to the meltdown.
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