Micromechanical behavior of UO2: crystalline anisotropy and associated internal stressezs in polycrystals

Abstract

Uranium dioxide is the standard nuclear fuel for pressurized water reactors in France. This ceramic is manufactured by sintering. The standard shape for use in nuclear reactor is a small cylinder, also called “pellet”, which measures about 8mm in diameter and 12mm in height. These pellets are then stacked into a zirconium alloy cladding, forming a rod. The fuel rods are then assembled together and these assemblies are put in the nuclear core. The mechanical behavior of the nuclear fuel during operation depends on the mechanical state of the rod (pellet and cladding), which can be related to other phenomena taking place during irradiation. A first step in the modeling approach is to study the mechanical behavior of non irradiated uranium oxide. For a temperature higher than about 1000°C and for a strain rate higher than about 1.10/6s/1, this material can be plastically deformed by a dislocation glide mechanism. For lower strain rate, the deformation mechanism is a diffusional process (see[6] for a complete map of the different deformations mechanisms in UO2). In this paper, we study the effect of the plastic anisotropy on the kinematic hardening of UO2. Our work is based on a microscopic approach and our results suggest that the kinematic hardening effect can be explained by the intergranular interaction between neighbor grains. An evaluation of this behavior was made using a crystal plasticity constitutive model and a polycrystalline aggregate.