Corrosion of zirconium fuel cladding inside a boiling water reactor: A post-irradiation study by atom probe tomography

The life-time-limiting factors of zirconium-based fuel cladding in water-cooled and -moderated nuclear power reactors are corrosion and associated hydrogen pickup. Corrosion performance in reactor is significantly worse in comparison to autoclave exposure. The accelerated degradation becomes particularly severe with the accumulation of radiation damage that is caused by fast neutrons. This work aims to expand the understanding of the underlying mechanisms governing the in-reactor corrosion process by nano-scale characterization of high-burnup fuel cladding tubes from operation in the boiling water reactor Oskarshamn 3 mainly by atom probe tomography. We present data from the oxide and the oxide-metal interface and point out the differences with the comparatively well-known behaviour in autoclave corrosion tests. The main aspect is the interaction between alloying elements, irradiation-induced defects and zirconium oxidation: Irradiation-induced FeCrNi clusters seem to slightly accelerate the diffusion of oxygen within the basal plane of the hexagonal metal matrix and dissolve in the oxygen-saturated zirconium metal that develops before zirconia formation takes place, and c-component dislocation loops, characteristic of high damage levels, might offer enhanced oxide nucleation sites that potentially explain the rapid degradation observed after some years of reactor operation. In addition, pores can be construed as a potential pathway for accelerated hydrogen pickup, similar to processes postulated in the literature. The results in this study give some novel insights into the mechanisms of in-reactor degradation of zirconium-based alloys and highlight the necessity to characterize materials from actual reactor operation.