Zirconium (Zr) alloys, serving as fuel cladding tubes and grids for pressurized water and boiling water nuclear reactors, undergo oxidation when exposed to oxidizing environment during service, directly impacting their operational lifespan. This review focuses on the oxidation behaviors of Zircaloys in two critical environments: waterside corrosion under normal service conditions and oxidation in high-temperature steam during accident conditions. We discuss the oxidation mechanisms and kinetics of Zr alloys, emphasizing how phase transformations in the zirconia (ZrO₂) scale influence the stability of the oxide film, thereby accelerating hydrogen uptake and failure processes. The oxidation behavior of Zr alloys is governed by complex factors, including alloy compositions, microstructures, and environmental conditions. This review aims to provide a comprehensive overview of the oxidation mechanisms involving ZrO₂ formation. It also explores computational methods for studying atomistic processes and discusses strategies to improve oxidation resistance. Finally, it outlines current research limits and future directions for developing accident-tolerant Zr alloys. This review discusses the oxidation mechanisms and kinetics of Zircaloys in nuclear reactors, focusing on waterside corrosion under normal service conditions, and oxidation in high-temperature steam during an accident.
Liu et al. (2026) studied this question.