EXPERIMENTAL INVESTIGATION OF HEAT TRANSFER COEFFICIENTS ON THE IPR-R1 TRIGA NUCLEAR REACTOR FUEL RODS

摘要

The IPR-R1 TRIGA Nuclear Research Reactor at Nuclear Technology Development Center (CDTN), Belo Horizonte (Brazil), is a pool type reactor cooled by natural circulation. Fuel to coolant heat transfer patterns must be evaluated as function of the reactor power in order to evaluate the thermal hydraulic performance of the core. The heat generated by nuclear fission is transferred from fuel elements to the cooling system through the fuel/cladding (gap) and the cladding to coolant interfaces. The fuel thermal conductivity of the fuel and the heat transfer coefficient from the cladding to the coolant were evaluated experimentally. A correlation for the gap conductance between the fuel and the cladding was also presented. In nuclear fuels the heat parameters become functions of the irradiation as a result of changes in its chemical and physical composition. As the reactor core power increases, the heat transfer regime from the fuel cladding to the coolant changes from single-phase natural convection to subcooled nucleate boiling. Results indicated that subcooled boiling occurs at the cladding surface in the central channels of the reactor core at power levels in excess of 265 kW.