Understanding the corrosion processes of fuel cladding in pressurized water reactors

The cladding made of zirconium alloy provides the first containment barrier for fission products, which is why its mechanical integrity is a prerequisite for nuclear safety. The external corrosion of the zirconium alloy cladding is one of the factors limiting the fuel rod’s lifetime. Understanding the corrosion processes of the zirconium alloys is consequently very important industrial issue for safety and efficiency of light water reactors. This review of international works is divided into two parts, dealing with the oxidation behavior and the hydrogen pickup (HPU) of zirconium alloys in pressurized water reactor environments. The first one describes the growth mechanism of the oxide, the oxidation kinetics and its modeling, the crystallographic phases and microstructure of the oxide layer, the alloying element effect and the irradiation impact on the corrosion rate. The second part is focused on the HPU mechanism and the link between oxidation and hydriding processes.