A model to describe the interaction between UO 2 and zircaloy in the temperature range 1000–1700°C

Abstract

During high temperature transients the Zircaloy claddings in a nuclear power reactor may chemically interact with their environment, either on the internal or on the external surfaces. The external oxidation due to the interaction with the coolant has been extensively studied and described in models. However, the internal oxide-metal interaction is less well known and has not yet been satisfactorily described in models. In this paper the internal oxidation, which results from the reduction of the UO 2 pellets by the Zircaloy cladding material is analyzed. A model is developed which solves only the oxygen diffusion problem in the five phases which are formed due to the UO 2/Zircaloy interaction, without taking into account zirconium and uranium diffusion. The oxygen diffusion coefficients of four of the phases are determined and the oxygen concentrations at some of the interfaces are adjusted in order that the interface movements be in accordance with the experimental data published in the literature for the temperature range 1000 to 1700°C. The model attempts only to simulate the kinetics of the system during the first stages of the process when all the interface movements obey parabolic rate laws.