Oxidation of zirconium alloys for nuclear fuel cladding

communications materials Review article A Nature Portfolio journal https://doi.org/10.1038/s43246-026-01201-1 Oxidation of zirconium alloys for nuclear fuel cladding Check for updates 1234567890():,; 1234567890():,; Tian-Yu Liu 1 & Wei-Zhong Han2 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 hightemperature 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. As a clean energy source characterized by high energy-conversion efficiency and low-carbon emissions, nuclear power maintains a strategically pivotal role in global energy transitions1. However, its safety and reliability remain primary constraints on development2. Zirconium (Zr) alloys are widely used in nuclear reactors due to their low neutron absorption, excellent corrosion resistance, and adequate mechanical properties3. The alloy fuel cladding is the primary barrier against the release of radioactive fission products and is exposed to synergistic degradation environments for long periods of time during reactor operation. These mechanisms include long-term exposure to high-temperature and high-pressure water environments, coupled with the influence of dissolved oxygen (DO), hydrogen (DH), and chemical additives, leading to waterside corrosion4. Under accident conditions, such as coolant loss, the temperatures can rapidly exceed 1000 °C, triggering severe steam oxidation. The synergistic effects of these service conditions accelerate the oxidation kinetics of the cladding material. The progressive accumulation of internal stress within the oxide layer promotes the formation of transverse cracks near the interface, which compromise the integrity of the oxide layer. These cracks also provide fast pathways for hydrogen diffusion and reduce the operational lifespan5. Zircaloys interact with aqueous media during both normal coolant service and high-temperature steam oxidation in accidents. This interaction forms zirconia (ZrO₂) and generating hydrogen species. The ZrO₂ layer formed on the cladding surface acts as a protective barrier against corrosive species and hydrogen ingress, thereby mitigating substrate corrosion4. Under normal coolant contact conditions (280–350 °C), a dense ZrO₂ forms on the surface, through which hydrogen slowly permeates via oxide defects, accumulating over time. Long-term operation and high burn-up conditions lead to thickening of the oxide layer, phase transformations, and development of defect oxides. However, the precise mechanisms governing these microstructural evolution pathways and their relative contributions to degradation remain a subject of active debate, causing a decrease in the integrity of the oxide film. In accident conditions, Zircaloys interacting with steam develop the oxide structure denser in the inner layer and porous in the outer layer. The oxidation behaviour of Zircaloys, particularly their oxygen interaction kinetics and ZrO2 formation thermodynamics, fundamentally determines nuclear fuel cladding service life. Therefore, understanding these oxidation mechanisms and defect evolution pathways is critical for predicting cladding degradation and optimizing accident-tolerant fuel designs. The early development of Zircaloys focused on enhancing mechanical strength and creep resistance through strategic additions of tin (Sn), iron (Fe), and chromium (Cr)6. As corrosion resistance grew in importance, the Soviet Union pioneered the integration of niobium (Nb), leading to the development of the Zr-1Nb alloy. Subsequent advancements involved compositional optimization of Sn, Nb, and trace elements (e.g., oxygen), yielding advanced cladding materials such as ZIRLO, E110, HANA-series, M5, and N18 alloys. Recent research explores adding trace elements like Yttrium (Y) and metalloids (e.g., Silicon (Si), Germanium (Ge)) to improve oxide film self-healing capacity and corrosion resistance7–9. However, the 1 Center for Advancing Materials Performance from the Nanoscale, State Key Laboratory for Mechanical Behaviour of Materials, Xi’an Jiaotong University, Xi’an, 710049, PR China. 2Department of Materials Science and Engineering, Department of Mechanical Engineering, City University of Hong Kong, Hong Kong e-mail: SAR, China. Communications Materials | (2026)7:137 1 Review article https://doi.org/10.1038/s43246-026-01201-1 Table 1 | Compositions of typical Zircaloys for nuclear fuel cladding6,10 Alloys Alloy composition (wt.%) Remarks Ni Sn Nb Cr Fe Others Zr-1 / 2.5 / / / / Not suitable Zr-2 0.05 1.5 / 0.1 0.12 0.12 O BWR Zr-4 / 1.5 / 0.1 0.2 0.09-0.13 O PWR E110 / / 0.95-1.05 / / ≤0.10 O PWR/RBMK/VVER E125 / / 2.20-2.60 / / 2.20-2.60 O PWR/RBMK/VVER E635 / 1.10-1.30 0.95-1.05 / 0.3-0.4 0.05-0.12 O PWR/RBMK/VVER ZIRLO / 1.02 1.01 / 0.1 0.09-0.15 O PWR OPT ZIRLO / 0.60-0.80 1.02 / 0.11 1.04 O PWR M5 / / 1 0.015 0.05 0.09-0.13 O PWR NDA 0.01 1 0.1 0.2 0.3 0.12 O PWR AXIOM X1 / 0.3 0.7-1 / 0.05 0.12Cu,0.2 V PWR HANA-4 / 0.4 1.5 / 0.1 0.12Cu,0.2 V PWR HANA-6 / / 1.1 / / 0.05Cu PWR N18 / 1.0 0.3 0.1 0.3 / PWR N36 / 1.0 1.0 / 0.3 / PWR CZ1 / 0.9-1.4 0.1-0.3 0.07-0.25 0.3-0.5 0.05-0.3Cu PWR efficacy and underlying mechanisms of these novel alloying additions are not fully established, with literature reports sometime presenting conflicting results regarding their optimal concentrations and long-term stability. Due to harsh service conditions, the oxidation resistance of Zircaloy cladding must meet more stringent standards. Extensive research has focused on the interaction between Zircaloys and corrosive environments, particularly their oxidation kinetics and mechanisms, which involve phase transformations during oxide growth and defect formation within the oxide layers. Nevertheless, critical knowledge gaps persist regarding the fundamental mechanisms governing these processes, and contradictions (...truncated)