Techniques for the Irradiation of Ceramic Fuels in a Moderate Thermal Neutron Flux

This article considers the physics calculations required to determine power, heat flux, and burnup for ceramic fuel materials as a function of variables such as fuel radius, enrichment, reactor thermal neutron flux, and irradiation time. I t is demonstrated t h a t high fuel burnups may be obtained using moderate thermal neutron flux by proper choice of variables. Heat transfer calculations utilizing the thermal resistance concept for a specific capsule design are described, together with an analysis of the design and operational uncertainties.