CFD analysis of natural convection heat transfer in volumetrically heated corium melt for CANDU reactors

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In March 2011, the Fukushima Daiichi nuclear power plant was subjected to an earthquake and tsunami resulting in a severe accident to Units 1, 2, and 3. During this accident molten fuel breached the reactor pressure vessel wall which resulted in a loss of containment of radioactive material. There was significant radiological release to the environment and over 100,000 people evacuated from the area. The Fukushima Daiichi disaster highlighted that in-vessel retention of molten fuel is of paramount importance. Under severe accident conditions, such as what occurred in Japan, the molten fuel and other internal reactor components form a liquid pool of corium melt at the bottom of the reactor vessel. Natural convection flows develop due to the volumetric heating within the melt. The volumetric decay heat from the fission process must be adequately removed at the vessel wall in order to ensure in-vessel retention of the molten material. Canada deuterium-uranium (CANDU) reactor vessels (for example at Darlington Nuclear Generating Station and Bruce Power Nuclear Generating Station) are surrounded by a cylindrical shield tank containing liquid water which acts to provide external vessel cooling in the event of a severe accident. Numerical analysis is used to assess the consequences of severe accidents and to determine if an adequate safety margin is in place to ensure that the reactor vessel integrity is maintained. In the nuclear industry, integral or lumped parameter codes primarily use empirical heat transfer correlations to predict the heat transfer from the molten corium pool to the vessel wall in order to assess if cooling of the exterior of the reactor vessel wall is sufficient to maintain its integrity. The proposed research will use computational fluid dynamics (CFD) to obtain detailed information on the natural convection fluid flow, and local (spatial) and time-varying wall heat transfer rates. These will be used to assess the appropriateness of underlying assumptions in the simplified integral codes which are currently used for beyond design basis analysis in the nuclear industry.