Analysis of natural convection heat transfer and solidification within a corium simulant

In the event of a severe nuclear accident such as that at the Fukushima Daiichi Nuclear Power Plant (NPP) in 2011, it is essential to minimize the release of radioactive material into the environment. Studies on Canada Deuterium Uranium (CANDU) reactors showed that thermal stresses are a potential threat to vessel wall integrity. There is thus a need to analyse the heat transfer and fluid flow within the corium and determine the spatial distribution of heat flux at the calandria vessel wall to ensure the ex-vessel cooling is sufficient. Canadian Nuclear Laboratories (CNL) designed and conducted an experiment to analyse the heat transfer within the corium using a 1/5 scale CANDU calandria vessel with molten salts as the corium simulant. The current study used computational fluid dynamics (CFD) to simulate the heat transfer and crust formation in the experiment to provide insight into the flow pattern and heat transfer and to assess the adequacy of the CFD modelling. Unsteady Reynolds-Averaged Navier–Stokes (URANS) and enthalpy-based solidification models were used to simulate the experiment. The numerical results show that crust forms along the vessel wall up to an approximate polar angle of 42 ° . There are natural convection circulations close to the top surface, and boundary layers form along the vessel wall. The results predicted by the numerical model match closely with the experimental results.