Experimental determination of heat transfer coefficients in uranium zirconium hydride fuel rod

Abstract

The heat generated by nuclear fission is transferred from fuel elements to the cooling system through the fuel- to- cladding gap and the cladding-to-coolant interfaces. The fuel thermal conductivity 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. 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 claddingsurface in the central channels of the reactor core at power levels in excess of 265 kW.