Irradiation cell for thermal-neutron fission of small samples 235U

Abstract

We have attempted to give an indication of the importance of heat transfer, during the thermal-neutron fission of small samples 97·46 per cent-enriched uranium. We irradiated two 280 mg 235U samples, in the UO 3 oxide powder form, in the central vertical port of the UFTR (University of Florida Training Reactor, Gainesville, Fla., U.S.A.) at 75 kW for 10 hr. Each sample was put into four quartz ampules, placed in aluminum containers and water cooled. Heat flow in the radial direction of the samples was calculated, assuming that: (1) the resistance to heat transfer in the contact areas between solid materials is not negligible, (2) the thermal conductivities of the fuel, the quartz, the aluminum and the air, as well as the physical properties of the coolant (density, viscosity, specific heat) are constant and independent of temperature. The errors involved in this assumption depend on the severity of the temperature gradients and the dependence of the above properties on temperature. This method is accurate if they are linear functions of temperature.