Role of ZrO2 oxide layer on the fretting wear resistance of a nuclear fuel rod

Abstract The ZrO2 oxide layers of Zr-based fuel claddings have been intensively studied to unveil their oxidation mechanisms in high temperature pressurized water. Nevertheless, there has been insufficient research on their mechanical properties, which are key factors determining the resistance to grid-to-rod fretting damage in normal Pressurized Water Reactor (PWR) operation. An experimental approach was applied to examine the tribological behavior of time-dependent oxide layers on both Zr cladding and grid, which were prepared in simulated PWR conditions for up to 360 days. It was found that the wear rate of pre-oxidized Zr cladding suddenly dropped with increasing oxide thickness of both the cladding and grid. The increase of surface roughness with oxide growth on the Zr-based grid could result in a rapid increase of wear damage by third-body abrasion. The well-developed columnar structure of the ZrO2 oxide layer could have a detrimental effect on the resistance to plastic deformation due to the enlarged grain size and relaxation of compressive residual stress by tetragonal to monoclinic ZrO2 transformation and crack formation. Consequently, ZrO2 oxide layers formed on fuel cladding and spacer grid under high temperature pressurized water show sufficient ductility to accommodate plastic deformation, which results in enhanced fretting wear resistance.