Oxidation of zirconium alloys for nuclear fuel cladding

Abstract

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.

Introduction

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 resistance^7,8,9. However, the 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 remain unresolved in current theories. Since oxidation behaviour critically determines cladding lifetime, a systematic review to synthesize prior studies and identify unresolved challenges is urgently needed. This review focuses on a comparative and correlative analysis of oxidation behaviour in both waterside and steam environments. It also discusses the influence of various factors on oxidation behaviour as reported in existing studies. Furthermore, it clarifies the key barriers to enhancing oxidation and corrosion resistance, and provides strategic guidance for extending cladding lifespan and designing next-generation high-performance Zircaloys.

Oxidation mechanism and kinetics

Several mature Zircaloy systems have been deployed in the nuclear energy sector, with Table 1 detailing the compositions of representative alloys^6,10. The specific development history of Zircaloys is illustrated in Fig. 1(a). Both waterside corrosion and accident steam oxidation involve oxide formation from reactions between Zircaloys and corrosive media. Figure 1(b) shows three processes by which Zircaloys react with water. First, water molecules from the coolant or steam physically adsorb onto the alloy surface and dissociate, releasing oxygen anions. Second, Zr atoms within the metal matrix lose electrons to form Zr cations (Zr⁴⁺). The subsequent oxide film thickening depends on the diffusion of oxygen anions. The ZrO₂ formed is an n-type semiconductor, wherein oxygen vacancies (${V}_{O}^{..}$) serve as the major charge carriers. These vacancies significantly promote oxygen ion migration via a vacancy mechanism. In contrast, the concentration of Zr vacancies (${V}_{Zr}^{{\prime} {\prime} {\prime} {\prime} }$) is extremely low, making Zr⁴⁺ cation diffusion negligible. Thus, the third stage of Zircaloy oxidation involves oxygen ions diffusing through the oxide film to the metal/oxide interface, where a reaction occurs. Oxidation nucleation preferentially occurs at surface defects or sites of elemental enrichment, leading to the formation of isolated ‘oxide islands’. As oxidation proceeds, these islands expand and coalesce, ultimately forming a continuous oxide film on the metal surface. Simulation methods such as first-principles calculations and molecular dynamics are employed to investigate atomic-scale oxidation mechanisms. These approaches calculate parameters of oxygen adsorption energy and atomic migration rates to quantify the oxidation process of Zircaloys^11,12. For a detailed discussion of the specific atomic mechanisms, see the section “Atomistic mechanism of oxidation in Zircaloys”.

Fig. 1: Oxidation behaviours of Zircaloy.

(a) Development history of Zircaloys for nuclear fuel cladding; (b) Oxidation mechanism of Zircaloys; (c) Schematic diagram of the oxidation kinetics of Zircaloys. Reproduced with permission from ref. ¹⁵, copyright (ASTM International, 2008); (d) Scanning electron microscope image of oxides formed on Zircaloys at 1000 °C with 100 hPa of steam partial pressure for 10000 s. Reproduced with permission from ref. ¹⁶, copyright (Elsevier, 2021); (e) Weight gains versus time during oxidation at 900 °C for 2 h of a M5 alloy exposed to air Reproduced with permission from ref. ²³, copyright (Elsevier, 2022); (f) Cross-sectional morphologies of M5 alloy oxidized in air at 900 °C for 80 min. Reproduced with permission from ref. ²³, copyright (Elsevier, 2022).

Table 1 Compositions of typical Zircaloys for nuclear fuel cladding^6,10

In waterside corrosion environments (280–350 °C, 10–16 MPa), Zircaloys show periodic oxidation kinetics¹³. Initially, the oxidation follows a parabolic rate law, as shown in Fig. 1(c)^14,15. Under prolonged exposure and cyclic loading, the kinetics shift to a near-linear regime with accelerated oxidation. This transition corresponds to oxide layer cracking (Fig. 1(d)). Localized stress concentrations exceed the fracture toughness of the ZrO₂ layer, causing cracks in the oxide film. The loss of oxide integrity provides fast diffusion paths for corrosive media¹⁶.

In steam environments, oxide film nucleation and growth govern metal oxidation kinetics. Higher temperatures accelerate oxidation in Zircaloys, yielding kinetics distinct from waterside corrosion^17,18. For instance, pure Zr in steam (415 °C, 0.2 MPa) oxidizes at a near-linear rate. Isotope tracing shows oxygen diffuses faster in pure Zr than in Zircaloy-4¹⁹. After extended exposure at 600 °C, Zr-1Nb alloy shifts from cubic to linear oxidation kinetics²⁰. In 1954, Jack et al.²¹ first systematically studied the high-temperature oxidation kinetics of low-hafnium Zircaloys in the range of 575–950 °C. By combining Wagner’s oxidation theory²² with experimental observations, they found that the initial mass gain behaviour deviated from classical parabolic kinetics, asymptotically approaching parabolic kinetics only after prolonged oxidation exposure.

Under extreme loss-of-coolant accident (LOCA) conditions (800–1200 °C), Zircaloy cladding heats rapidly and oxidizes quickly in steam. After oxidation at 900 °C for 80 min, the M5 alloy’s oxidation weight‑gain curve shows a clear inflection point and acceleration (Fig. 1(e)²³). Localized thick oxide areas, called “blisters” or “bulges,” appear on the surface (Fig. 1(f)²³), reflecting nodular corrosion. This non‑uniform oxide growth is typical in steam environments.

The content of alloying elements also influences the oxidation kinetics of Zircaloys, with the oxidation rate of Zr-Nb alloys increasing with higher Nb content. After oxidation in steam within the temperature range of 700–1200 °C, the Zr-1.0Nb-1.0Sn-0.1Fe alloy exhibited a faster oxidation rate compared to Zircaloy-4. Both alloys primarily followed parabolic oxidation kinetics across most temperature ranges. However, the oxidation weight gain patterns of both alloys deviate from the parabolic curve between 800 °C and 1050 °C. The deviation at 800 °C stems from phase transformations within the metallic matrix, while lateral cracks were observed in the oxide layer at both 1000 °C and 1050 °C²⁴. The reaction of Zircaloys with water or steam is accompanied by hydrogen generation. Under normal service conditions, hydrogen produced by the reaction gradually penetrates into the matrix, leading to hydrogen uptake in Zircaloys. Studies have revealed a correlation between the oxidation kinetic exponent of Zircaloys and their hydrogen absorption fraction in pure water at 18.7 MPa and 360 °C²⁵. Therefore, understanding oxidation mechanisms and kinetics is crucial for designing Zircaloys with better oxidation resistance and lower susceptibility to hydrogen embrittlement. Nevertheless, quantitatively predicting the hydrogen uptake fraction based on oxidation kinetics alone remains difficult, indicating persistent gaps in our fundamental understanding of the coupled oxidation-hydrogenation process.

Phase structures and defect generation in oxide scales

The protective effect of the oxide scale, which serves as the primary barrier against corrosion, depends directly on its structural integrity and phase composition²⁶. Elucidating the microstructural evolution and phase transformations in ZrO₂ scales is critical for mitigating hydrogen uptake and enhancing oxidation resistance in Zircaloys. Researchers have demonstrated that the oxide scale microstructure significantly influences the oxidation resistance of Zircaloys. As shown in Fig. 2(a), according to the Zr-O phase diagram¹⁰, different Zr-O phases form depending on the oxygen content. In a sufficient oxygen atmosphere, the oxide film of Zircaloys primarily consists of ZrO₂, with its main phases being the tetragonal phase (t-ZrO₂) and the monoclinic phase (m-ZrO₂). A metastable transitional phase of cubic-ZrO (c-ZrO) and hexagonal-ZrO (h-ZrO) also exists during the oxide formation process^27,28. The t-ZrO₂ phase is thermodynamically stable above 1200 °C, while the m-ZrO₂ becomes the thermodynamically stable phase at lower temperatures. A transformation between these two crystal structures occurs under high-temperature conditions^29,30,31,32. However, under high-stress conditions, the t-ZrO₂ phase can persist as a metastable phase at low temperatures.

Fig. 2: Phase transformation of oxides in Zircaloy.

(a) Zr-O binary phase diagram. Reproduced with permission from ref. ¹⁰, copyright (Elsevier, 2012). (b) Two typical Zr dioxide phases and their transformation. (c) Variation of the average stress in the oxide layer of Zircaloy with oxide thickness. Reproduced with permission from ref. ²⁶, copyright (Elsevier, 2015). (d) Schematic showing the key parameters affecting the characteristics of zirconia Reproduced with permission from ref. ⁴², copyright (Elsevier, 2020).

In a high-temperature and high-pressure water environment, metastable t-ZrO₂ is initially formed during the oxidation of Zircaloys, which subsequently gradually transforms into the m-ZrO₂ under thermodynamic driving forces and stress induction, as illustrated in Fig. 2(b). This phase transformation involves 3$\sim$5% volumetric expansion, which triggers microcracking, interfacial delamination, and morphological bifurcation in the oxide scale³³. The Pilling-Bedworth ratio (PBR) for ZrO₂ on Zr is 1.56, representing the oxide-to-metal molar volume ratio³⁴, leading to compressive stress accumulation within the oxide scale due to volumetric expansion during oxidation. This stress evolution demonstrates thickness-dependent behavior, as shown in Fig. 2(c)²⁶, where the residual stress in Zircaloy oxide layers first increases and then decreases with progressive oxide growth. Experimental measurements across various oxidation environments reveal an average stress range of 0.5–3 GPa for Zr alloys^{35,36,37,38,39}. It is widely recognized that localized stress concentrations and residual stresses within the oxide scale play a pivotal role in governing the zirconia phase transformation and subsequent degradation mechanisms. High-stress areas near the oxide/metal interface retain more t-ZrO₂. In contrast, regions farther from the interface experience stress relaxation through oxide cracking or metal creep. This relaxation triggers the transformation from t-ZrO₂ to m-ZrO₂^40,41, as shown in Fig. 2(d)⁴².

The oxide film on oxidized Zircaloy exhibits a curved morphology at both its surface and the interface with the metal, as shown in Fig. 3(a)⁴³. The m-ZrO₂ in the oxide film can be revealed through high-resolution transmission electron microscopy (HR-TEM), as shown in Fig. 3(b, c)⁴³. It is widely recognized that the volume fraction of the t-ZrO₂ progressively decreases with prolonged oxidation time, while the t-ZrO₂ phase predominantly resides near the metal/oxide interface, as shown in Fig. 3(d)⁴⁴. Atom probe tomography (APT) measurements have revealed that the Zr/O ratio at the interface exceeds 0.5. This significant deviation from the stoichiometric ratio may stem from the aggregation of point defects in amorphous zirconia or the formation of precursors for metastable zirconia (such as ZrO) under extreme conditions, as illustrated in Fig. 3(e)⁴⁵. Furthermore, the oxide film on ultrasonically surface-rolled Zr alloys contains a detectable fraction of t-ZrO₂, which is located predominantly near the metal/oxide interface, as shown in Fig. 3(f) and (g)⁴⁶.

Fig. 3: Microstructures of oxides in Zircaloy.

(a) Transmission electron microscope (TEM) image of oxides in Zircaloy-4 alloy after hydrothermal corrosion at 360 °C /18.7 MPa. Reproduced with permission from ref. ⁴³, copyright (Elsevier, 2023); (b) and (c) High-resolution TEM (HR-TEM) image of ZrO₂. Reproduced with permission from ref. ⁴³, copyright (Elsevier, 2023); (d) HR-TEM image of metal/oxide interface of a recrystallized (RXA) ZIRLO alloy corroded at 360 °C /18 MPa for 140 days. Monoclinic or tetragonal ZrO₂ grains are directly grown on α-Zr. Reproduced with permission from ref. ⁴⁴, copyright (Elsevier, 2012); (e) 10 nm slice from an atom probe tomography reconstruction showing the different oxide phases in Zircaloy-4. Reproduced with permission from ref. ⁴⁵, copyright (Elsevier, 2013); (f) TEM image of surface oxide layer on Zircaloy after ultrasonic surface treatment at a rolling speed of 200 r/min and corroded in water at 360 °C /18 MPa for 350 days. Reproduced with permission from ref. ⁴⁶, copyright (Elsevier, 2024); (g) HR-TEM image of interface region of the oxide film in (f). Reproduced with permission from ref. ⁴⁶, copyright (Elsevier, 2024).

Corrosion of α-Zr in pure water at 360 °C and 18 MPa leads to the formation of a novel hexagonal-ZrO (h-ZrO) phase at the oxide/metal interface, accompanied by distinct crystallographic orientation relationships, as shown in Fig. 4(a–c)²⁸. The transformation of zirconia phase is influenced by both stress and alloying elements. Another identified metastable phase, cubic-ZrO (c-ZrO), is also found predominantly near this interface²⁷. Studies have confirmed that this suboxide layer can enhance corrosion resistance after the transition⁹. Excessive Nb addition in Zr-Nb alloys reduces the volume ratio of t-ZrO₂/m-ZrO₂ phases and corresponding compressive stresses in the oxide scale²⁰. Furthermore, phase transformations in the oxide scale are correlated with coolant chemistry. For example, lowering the dissolved hydrogen concentration in high-temperature water decreases t-ZrO₂ content and increases m-ZrO₂ in Zr-Nb-Cu alloys due to stress-assisted transformation⁴⁷. Thermodynamic modelling indicates that t-ZrO₂ content near the interface, interfacial compressive stresses, stress gradients, and internal stresses from the phase transformation critically control microstructural changes. These include the transition from columnar to equiaxed grains, crack nucleation, and degradation of oxidation resistance⁴⁸. The factors influencing the tetragonal-to-monoclinic phase transformation in zirconia are complex, and current mechanistic explanations for the phase transition in oxide films remain limited.

Fig. 4: Characteristics of oxides in Zircaloy.

(a) Band contrast map showing oxide/Zr interface in Zr-0.5 Nb; (b) Phase map and (c) Texture component maps for m-ZrO₂, h-ZrO and α-Zr. Reproduced with permission from ref. ²⁸, copyright (Elsevier, 2019); (d) Proportion of t-ZrO₂ forming on Zircaloy-4 oxidized in the temperatures range of 800 °C to 1100 °C. The vertical axis T represents t-ZrO₂, while M represents m-ZrO₂. Reproduced with permission from ref. ²⁹, copyright (Elsevier, 2015). (e) Bright-field TEM image of oxide formed on Zircaloy-4 after oxidized in deionized water for 45 days at 360 °C. Reproduced with permission from ref. ⁵⁸, copyright (Elsevier, 2022); (f) (0$111$)-orientation map of Zr substrate grains of (e), the black curve shows the variation of pore density. Reproduced with permission from ref. ⁵⁸, copyright (Elsevier, 2022); (g) Cross-sectional bright-field TEM micrographs obtained from the oxide layers in Zircaloy-4, the inset shows the enlarged TEM image of a grain boundary containing a nano-pore. Reproduced with permission from ref. ⁵⁷, copyright (Elsevier, 2010); (h) Grazing incidence X-ray diffraction spectroscopy (GIXRD) patterns and (i) Raman spectra of the pre-oxidized Zircaloy-4 before and after irradiation. Reproduced with permission from ref. ⁶¹, copyright (Elsevier, 2023).

In steam environments, particularly between 800–1200 °C, the phase structure of surface ZrO₂ differs from that formed in low-temperature water, as temperatures approach the t-ZrO₂ to m-ZrO₂ transformation threshold. High-temperature steam environments induce porosity within the zirconia layer⁴⁹. Current research on the phase structure of surface zirconia after steam oxidation primarily focuses on post-experiment analyses conducted after cooling to room temperature⁵⁰. Notably, Gosset et al. performed in-situ X-ray diffraction analysis of the phase structure of Zircaloy surface oxide layers in high-temperature steam environments (800–1100 °C). As shown in the Fig. 4(d), when the temperature exceeds 1100 °C, the surface of Zircaloy-4 is almost entirely composed of the t-ZrO₂ phase. At oxidation temperatures below 1050 °C, the ZrO₂ formed during oxidation is biphasic. Notably, while the fraction of t-ZrO₂ decreases at lower oxidation temperatures, a higher fraction is paradoxically retained upon cooling to room temperature. During gradual cooling, a portion of t-ZrO₂ phase transforms into m-ZrO₂ phase, and this cooling process causes irreversible changes to both the grain size and the phase fraction of the ZrO₂²⁹.

The phase transformation in Zircaloys directly affects the structural integrity of their oxide films. Most researchers agree that zirconia films with a higher t-ZrO₂ content demonstrate better oxidation resistance^51,52,53. The transformation from t-ZrO₂ to m-ZrO₂ is accompanied by volume expansion and shear strain, inducing micron-scale cracks and interfacial delamination within the oxide film. These defects not only provide rapid diffusion pathways for oxygen ions and hydrogen, accelerating substrate corrosion and hydrogen embrittlement, but also significantly degrade the creep resistance and fatigue resistance of the oxide film^54,55. Research on the breakaway acceleration during high-temperature oxidation attributes this phenomenon to the t-ZrO₂ to m-ZrO₂ transformation⁵⁶.The phase transition induces transverse and radial cracks in the oxide film, thereby accelerating oxidation degradation, as shown in Fig. 4(e–g)^57,58. Studies have also observed nanopores and nanotubes within the zirconia film⁵⁹. The outer oxide layer contains a high density of nanotubes, while near the oxide/metal interface, continuous nanotubes fragment into isolated nanopores. The presence of these nanopores and nanotubes facilitates hydrogen diffusion into the substrate, ultimately triggering hydrogen embrittlement. Ni et al.⁶⁰ utilized focused ion beam (FIB) sectioning and 3D reconstruction to conduct quantitative analysis of crack microstructures and metal/oxide interface morphology in oxides formed on ZIRLO alloy under high-temperature conditions. The results revealed that cracks and nanopores within the oxide film establish direct pathways between the alloy substrate and oxidizing environment, significantly accelerating oxidation rates through enhanced ingress of oxidizing species into the metallic matrix.

The cracks and pores induced by phase transformation in oxide films are inevitable, but elemental additions can modulate their formation. In Zr-Nb-Y alloys, nanoscale pores originate from yttrium (Y) segregation at columnar oxide grain boundaries and the subsequent inhomogeneous interdiffusion of Y and Zr cations within the oxide lattice. This mechanism has been experimentally validated in M5 and N18 Zircaloys⁷. The phase transformation of zirconia from t-ZrO₂ to m-ZrO₂ on Zr alloy surface induces significant stress accumulation, leading to defect formation. Therefore, suppressing or delaying the tetragonal-to-monoclinic phase transition in zirconia can effectively reduce defects in the oxide layer, thereby enhancing the oxidation resistance of Zircaloys. Ion irradiation experiments on pre-oxidized Zircaloy-4 have shown that irradiation damage can induce a phase transformation from m-ZrO₂ back to t-ZrO₂, as shown in Fig. 4(h, i)⁶¹. This phenomenon is attributed to the compressive stress induced by neutron bombardment in-reactor, which effectively delayed the tetragonal-to-monoclinic phase transformation. Additionally, strategies such as modifying grain size, engineering stress states, and optimizing alloy composition can effectively suppress the tetragonal-to-monoclinic phase transformation in zirconia^54,62.

Critical factors impacting oxidation behaviour

The oxidation behaviour of Zircaloys is affected by a variety of factors, among which the key factors include alloy composition, microstructure, oxidation environment, and multiphysics coupling in reactor cores. The investigation of the factors affecting the oxidation behaviour of Zircaloys is of great significance in mitigating the failure of Zr alloy cladding. The influence of these factors on the oxidation behaviour of Zircaloys will be discussed separately in the following paragraphs.

Alloy composition

The primary alloying elements in Zircaloys are Sn and Nb, with minor additions including Fe, Cr, and O. In recent years, trace elements such as Cu and Si have been explored to further optimize the self-healing capacity and corrosion resistance of the oxide films^{63,64,65,66,67}. Therefore, the existing Zr alloys primarily consist of Zr-Sn, Zr-Nb, and Zr-Sn-Nb systems. The addition of these elements significantly influences the oxidation resistance of the alloys, as shown in Fig. 5(b)⁵⁹. The effect of alloying elements on the oxidation resistance also indirectly affects the hydrogen absorption of the Zircaloys.

Fig. 5: Effect of water chemistry on oxides in Zircaloy.

(a) {103} pole figures of monoclinic ZrO₂ in recrystallized annealed (RXA) and stress-relieved annealed (SRA) Zircaloy-4. Reproduced with permission from ref. ⁸⁷, copyright (Elsevier, 2024); (b) Weight gain and hydrogen pickup curves for three selected corroded Zr alloys. Reproduced with permission from ref. ⁵⁹, copyright (Elsevier, 2019); (c) Weight changes of three alloys after exposed to different water chemistry at 330 °C and 14 MPa for 200 h. LiBW stands for Li/B water and PW is pure water. Reproduced with permission from ref. ⁷⁶, copyright (Elsevier, 2023); (d) Weight gain curves for RXA and SRA Zircaloy-4 tubes exposed to 360 °C water containing 1.5 ppm Li. Reproduced with permission from ref. ⁸⁷, copyright (Elsevier, 2024); (e) Effect of Li and B concentrations on oxide film thickness in Zircaloy-4. Reproduced with permission from ref. ⁹¹, copyright (Elsevier, 2012).

The addition of Sn was initially intended to mitigate the adverse effects of nitrogen⁶⁸. It is now widely recognized that reducing Sn content enhances the oxidation resistance of Zircaloys^69,70, Once the concentration of Sn higher than 1.5 wt.%, it starts to deteriorate the corrosion resistance of Zircaloys. Therefore, the Sn content is now reduced to the amount just necessary to counteract the N levels. Several low-Sn Zircaloys was developed based on this idea, such as ZIRLO and Zr-2.5Nb^6,71. The influence of Sn on the oxidation behaviour of Zircaloys also depends on the corrosive environment. For high-temperature and high-pressure water environments, Zr-0Sn alloys exhibit superior oxidation resistance. Nevertheless, the addition of Sn causes an earlier transition in the corrosion acceleration point of Zircaloys and promotes crack formation in the oxide layer. More Sn content also accelerates the transformation of ZrO₂ from tetragonal phase to monoclinic phase, which facilitate crack initiation and deteriorate the oxidation resistance of Zr alloys^69,72. For high-temperature steam environments, nanopores tend to form around Sn-containing secondary phase particles within the oxide layer, which can act as fast diffusion channel for corrosive media⁷³. The only except is that the Sn improves the oxidation resistance of Zircaloys in corrosive environments with LiOH-containing coolants. For instance, Zr-1Sn alloy demonstrates better corrosion resistance than Zr-0.2Sn alloy. Because Sn addition hinders the penetration of Li⁺ into the oxide layer, allowing the columnar oxide structure to remain intact and reducing the corrosion rate⁷⁴.

Nb as a secondary principal alloying element in Zr alloys was primarily introduced to enhance their corrosion resistance⁶³. It promotes the formation of fine β-Nb particles, which inhibit oxygen diffusion along grain boundaries and delay Zr oxidation, thereby stabilizing the oxide film⁷². However, excessive Nb addition increases the susceptibility of Zr alloys to oxygenated environments, ultimately inducing nodular corrosion in high-oxygen aqueous conditions⁷⁵, as shown in Fig. 5(c)⁷⁶.

Fe typically exists in Zircaloys as secondary phase particles (SPPs), such as Zr(Fe, Cr)₂ and Zr₂(Fe, Ni) intermetallic compounds, which serve as the primary pathway through which Fe influences the oxidation behaviour. It is revealed that the addition of Fe exhibits a dual influence on the corrosion resistance of Zircaloys. In Zr-1Nb alloy, the 0.4 wt.% Fe addition enhances corrosion resistance⁶⁵. However, corrosion resistance degrades when the Fe content reaches 0.45 wt.%⁷⁷. This decline is attributed to the increased population of SPPs at higher Fe levels. These SPPs oxidize slower than the surrounding Zr matrix. The oxidation mismatch induces severe stress concentration at the particles, ultimately causing the protective oxide film to crack at these sites⁷⁷.

Numerous studies have shown that the addition of Cr is beneficial for the corrosion resistance of Zircaloys. The combined effect of Cr and Fe improves the alloy’s resistance to nodular corrosion in high-temperature steam⁷⁸. Furthermore, research has shown that the incorporation of Cr into Zircaloys can reduce the hydrogen absorption rate and promote the formation of columnar crystals in the oxide film⁷⁹. In contrast, the introduction of Ni into Zircaloys tends to promote hydrogen absorption, increasing the risk of hydrogen embrittlement⁷⁷.

Cu significantly enhances the corrosion resistance of Zr-1Nb alloys in various environments. However, when the Cu concentration exceeds 0.2 wt.%, coarse secondary phase particles precipitate within the alloy, detrimentally compromising its corrosion resistance. In Y-modified Zr-1.0NbxY alloys, nanoscale pores form at the grain boundaries of columnar ZrO₂⁸⁰. This phenomenon originates from the accelerated diffusion of Y³⁺ compared to Zr⁴⁺ in the oxide lattice. According to the Kirkendall effect, vacancy fluxes generated by such cationic mobility differentials accumulate at grain boundaries, ultimately coalescing into nanopores⁷. Consequently, strategic optimization of alloying elements and their concentrations emerges as an effective pathway to improve oxidation resistance in Zircaloys.

Microstructures

The formation of the surface oxide layer on Zircaloy is closely related to the microstructure of the underlying metal matrix^81,82. Geng et al.⁴³ introduced a coarse-grained structure with <10$\bar{1}$0> texture on the surface of Zircaloy-4. After one hour of corrosion in 700 °C steam, the coarse-grained Zircaloy-4 exhibited a 40% reduction in weight gain compared to the initial alloy. The decreased grain boundary density in the coarse-grained structure reduced oxygen diffusion pathways, thereby enhancing oxidation resistance. However, when grain sizes fall below 100 nm (ultrafine grains), the material’s properties and microstructural characteristics differ significantly from those of coarse-grained counterparts⁴⁶. A computational model by Zhang et al.⁶² shows the oxidation rate constant decreases with grain size, enhancing oxidation resistance in nanocrystalline materials. Beyond grain size effects, low-solubility alloying elements such as Nb, Cr, and Fe exist as SPPs in Zircaloys, whose composition and distribution critically influence oxidation behaviour⁸³. As oxidation progresses, the metal/oxide interface migrates inward, and delayed oxidation of secondary phases relative to the matrix generates cracks around these phases due to oxidation-induced stresses⁸⁴. These cracks serve as preferential pathways for corrosive medium diffusion. Zhao et al.⁸⁵ demonstrated that interconnected microcracks around secondary phases form percolation networks in the oxide layer, altering the transport of hydrogen and oxygen ions.

The matrix texture also affects oxidation behaviour⁸⁶. Lin et al.⁸⁷ compared the corrosion performance of recrystallized annealed (RXA) and stress-relieved annealed (SRA) Zircaloy-4 alloys in lithiated water at 360 °C for 400 days. RXA Zircaloy-4, with stronger {0002} basal texture, showed significantly slower oxidation rates than SRA material, as shown in Fig. 5(a), (d). High dislocation densities in the matrix enhance oxygen pipe diffusion and promote localized oxidation initiation. However, Garner¹³ observed identical oxide microstructures on ZIRLO alloy surfaces with different matrix orientations after oxidation in 360 °C water, suggesting limited correlation between oxide orientation and substrate texture. While extensive studies have explored the influence of Zircaloy microstructure on oxidation behaviour, mechanistic explanations remain contentious, with unresolved discrepancies among researchers.

Oxidation environments

Current research on Zircaloy oxidation focuses mainly on temperatures between 300 °C and 1200 °C^23,88,89. The primary focus lies on the normal coolant service environment within high-pressure autoclaves, where Zircaloys are exposed to deionized water and lithiated water at 280–350 °C with pressure of 10–16 MPa. The reaction process is subject to synergistic degradation caused by high temperature, high pressure, dissolved hydrogen, oxygen and irradiation damage. The lithiated water environment is designed to control neutron reactivity by introducing boric acid, which exhibits an exceptionally high thermal neutron absorption cross-section, effectively capturing neutrons to suppress nuclear fission chain reactions. Adjusting boric acid concentration enables dynamic regulation of reactor power output. However, to prevent excessive boric acid addition from destabilizing the aqueous environment, lithium hydroxide (LiOH) is concurrently incorporated to maintain coolant pH. This chemically complex water environment significantly impacts the oxidation behaviour of Zircaloys⁹⁰.

Researchers have demonstrated that the interaction energy between LiOH and Zr surfaces exceeds that of boric acid, indicating that LiOH exerts a more pronounced influence on fuel cladding⁹⁰, as shown in Fig. 5(e)⁹¹. The addition of trace boron to LiOH effectively reduces weight gain in Zr-Sn-Nb alloys⁹². Ding et al.⁹³ exposed Zr-2.5 Nb pressure tubes to pressurized lithiated water environments with varying LiOH concentrations, identifying a critical LiOH concentration beyond which corrosion accelerates. At low LiOH concentrations, the surface zirconia layer acts as a protective barrier against Li⁺ diffusion⁹⁴. Under high LiOH concentrations, oxidation products of β-Nb phases in Zr alloy react with LiOH, leading to dissolution and subsequent porosity formation within the oxide film^95,96. In pure water environments, elevated Sn content accelerates the transition of oxide crystal structures from tetragonal to monoclinic phase. However, in LiOH solutions, increasing Sn content in the alloy conversely inhibits lithium ion transport through the oxide layer⁷⁴.

Furthermore, the concentrations of dissolved oxygen (DO) and dissolved hydrogen (DH) in the coolant significantly influence the oxidation behaviour of Zircaloys. Under hydrogenated, oxygen-depleted conditions, uniform corrosion occurs, whereas in non-hydrogenated, oxygen-containing environments, nodular corrosion is observed, predominantly in boiling water reactors (BWRs)⁹⁷. In water vapor environments with DO concentrations of 1000 μg/L, oxygen significantly influences the phase transformation of zirconia on Zircaloy surfaces, accelerating oxidation⁵⁵. However, reducing Sn content in the alloy can enhance its oxidation resistance under such oxygen-rich conditions⁹⁸. When temperatures exceed 1000 °C, the coolant rapidly transitions to steam-termed “accident conditions” in practical service. Severe oxidation of Zr alloy cladding by steam leads to massive hydrogen and heat release, ultimately compromising the integrity of the fuel barrier⁹⁹. Thus, the service environment is a critical factor governing the oxidation of Zircaloys.

Hydrogen absorption factors

During service, Zircaloys can absorb hydrogen from water reactions or the corrosive environment. Research shows that in cooling water, higher hydrogen concentrations increase the corrosion rate of Zircaloy-4 and accelerate its kinetic transition¹⁰⁰. Kido¹⁰¹ reported hydrogen-induced accelerated uniform corrosion in 360 °C water, while Blat et al.¹⁰² observed a similar phenomenon that samples with higher hydride content exhibited accelerated corrosion. When hydrogen exceeds its solubility limit, hydrides start to form in Zr alloys. High stress at the oxide/metal interface and low temperatures promote hydrogen segregation toward the interface, where most hydrides are formed¹⁰³.

Numerous studies have indicated that hydrides formed within the matrix can accelerate the oxidation of Zircaloys^104,105. Researchers conducted pre-hydriding treatment on Zircaloy-4 prior to oxidation and found that the oxidation rate increased with higher initial hydrogen content. The oxide layer formed on the matrix surface containing hydrides was thicker than that formed on a hydride-free matrix surface. This phenomenon is attributed to the influence of hydrides on the matrix lattice, which reduces the coherency between the matrix and the zirconia. As a result, the protective quality of the formed zirconia is compromised, leading to an accelerated oxidation¹⁰¹.

Multiphysics coupling in reactor cores

During actual operation, a nuclear reactor is continuously exposed to multiple physical fields, including heat transfer and neutron irradiation. In such an environment, the fuel cladding tubes undergo oxidation due to the interaction of these multiphysical fields. Serving as the primary barrier between nuclear fuel and coolant, fuel cladding not only prevents direct contact between fuel and cooling water but also facilitates heat transfer. The heat flux, defined as the thermal power released per unit area from the cladding surface to the coolant, serves as a critical parameter for evaluating the heat transfer efficiency between the fuel and coolant. During the LOCA, complete coolant loss prevents effective heat removal from fuel rods. The heat transfer efficiency between the cladding surface and the steam environment decreases extremely, causing the heat flux to approach zero. When the accumulated temperature in Zr alloy cladding exceeds 800–1200 °C, Zr reacts violently with steam, accelerating oxidation¹⁰⁶.

As fuel cladding, Zircaloys inevitably experience neutron irradiation during in-reactor service. Studies have revealed that neutron flux significantly impacts oxide formation on Zr surfaces, including higher defect concentrations in the material, redistribution of alloying elements, and oxide phase transformations. Early research suggested that neutron-induced defects accelerated oxidation by enhancing oxygen diffusion through the oxide layer. However, subsequent studies have challenged this view¹⁰⁷. The primary culprit for accelerated oxidation in reactors has been specifically identified as the redistribution of alloying elements^108,109. As previously mentioned, alloying elements play crucial roles in stabilizing oxide growth. The four principal elements of interest are Fe, Cr, Sn, and Nb. Due to the low solubility of Fe and Cr in Zr, they typically form Zr(Fe, Cr)₂ second phase in the matrix¹¹⁰. Nb forms a β-Nb solid solution phase in the matrix and also combines with trace Fe to form the Zr(Fe, Nb)₂ phase⁹⁷. Sn generally dissolves in the Zr matrix.

The studies show that Sn- containing alloys and Fe-containing alloys are more susceptible to radiation-accelerated corrosion than Nb-containing alloys^108,109. The redistribution of iron and dissolution of precipitates are considered responsible for the failure of anodic protection¹⁰⁹. In Nb-containing alloys, Nb precipitates exhibit minimal radiation-induced changes, showing only slight redistribution and forming ultra-fine needle-like Nb precipitates^109,111,112. Higher Sn content correlates with exacerbated out-of-reactor corrosion, complicating the assessment of its specific radiation effects. Nevertheless, Sn redistribution under irradiation has been observed. Brendan’s¹⁰⁰ study on Zircaloy-4 within 272–355 °C under varying neutron fluxes demonstrated that increasing neutron fluence enlarges oxide grain size, reducing the tetragonal phase fraction in oxide regions distant from the metal/oxide interface. Conversely, greater tetragonal phase content was observed at the metal/oxide interface with higher fluence. The combined effects of irradiation-induced elemental redistribution and gradual loss of metal ductility create unstable oxide growth. Elemental inhomogeneity induces stress accumulation that ultimately disrupts the protective oxide layer. Concurrent metal hardening exacerbates this process by impeding stress relief during oxide growth.

During prolonged reactor operation, oxidation products primarily composed of soluble metal cations and oxide particles are released into the coolant. These products are transported to the vicinity of the cladding tubes through coolant circulation. Under subcooled nucleate boiling conditions, these corrosion products accumulate on the fuel cladding tube surfaces, forming chalk river unidentified deposits (CRUD)¹¹³. The primary components include metal oxides such as iron oxide (Fe₃O₄, Fe₂O₃,), as well as boric acid added to PWR coolant that precipitates in localized areas¹¹⁴. The interaction between CRUD and the underlying oxide is complex. During the initial deposition stage, the CRUD layer acts as a physical barrier that inhibits the diffusion of corrosive species^115,116. In later stages, however, CRUD primarily accelerates oxidation. This shift occurs because the CRUD layer on the oxide surface impedes heat transfer, leading to localized temperature elevation at the CRUD/oxide interface and thereby enhancing cladding corrosion^115,116.

Under real operating conditions, the oxidation of Zircaloys is governed by the coupled effects of these multi physics fields, making the overall oxidation mechanism highly complex.

Atomistic mechanism of oxidation in Zircaloys

Oxidation reactions involve atomic or molecular-scale processes such as adsorption, diffusion, and phase transformations. The interaction between oxygen and metal surfaces is highly complex, and the oxidation behaviour of Zircaloys in their operating environments is challenging to monitor in real time. Existing characterization techniques struggle to directly observe oxygen atom adsorption and diffusion on metal surfaces. However, computational simulations can visually reveal oxygen migration pathways across metal surfaces and the growth dynamics of oxide layers. Therefore, simulation-based approaches for studying the oxidation behaviour of alloys have gained increasing application in recent years^11,12. Existing simulation methods for studying metal oxidation behaviour primarily include first-principles calculations (DFT), molecular dynamics (MD) simulations, phase-field modelling, and finite element analysis (FEA). Research on oxidation behaviour employs different simulation approaches at varying scales to investigate distinct aspects of the process. First-principles calculations based on DFT can resolve the interactions between oxygen atoms and metals, elucidating the driving forces and energetics of oxidation reactions^117,118. Lindgren¹¹⁹ employed first-principles modelling combined with experimental data to investigate cathodic and anodic reactions during Zr corrosion in water, as illustrated in Fig. 6(a) and (b), and identified charge-dependent oxygen vacancy migration in ZrO₂.

Fig. 6: Atomistic structures of oxides in Zircaloy.

(a) Different views of a Zr₆O octahedral site in Zr with 1.5 at. % O. Reproduced with permission from ref. ¹¹⁹, copyright (Royal Society of Chemistry, 2014). (b) Chains of edge-sharing octahedra, diffusion path and corresponding energy profiles of Zr with 20 at. % O. Reproduced with permission from ref. ¹¹⁹, copyright (Royal Society of Chemistry, 2014). (c) Variation of energy for O atom diffuses from H3 site to adjacent H2 site along [010] orientation. Reproduced with permission from ref. ¹²⁰, copyright (Elsevier, 2021). (d) Variation of energy for O atom diffuses from H2 site to adjacent B1 site. Reproduced with permission from ref. ¹²⁰, copyright (Elsevier, 2021). (e) Molecular dynamics (MD) simulations of water, OH^–, Li⁺ and boric acid. (f) Pure water models obtained by the amorphous cell module. (g) Alloy-LiBW layered model after geometry optimization for MD simulations. Reproduced with permission from ref. ⁷⁶, copyright (Elsevier, 2023).

Additionally, first-principles calculations can be used to probe the influence of substrate characteristics on oxidation processes. Liu et al.¹²⁰ employed first-principles calculations to study the interaction between oxygen and the (101$\bar{1}$) surface of α-Zr. As shown in Fig. 6(c)and (d), the energy of oxygen atoms varies across different adsorption sites on the (101$\bar{1}$) surface, with oxygen ultimately prefers to adsorb at the lowest-energy site. However, first-principles simulations primarily focus on ground or transition states and struggle to model large systems or long-timescale dynamic processes.

Molecular dynamics simulations can track characteristics such as oxide film growth and grain boundary diffusion, as shown in Fig. 6(e-g)⁷⁶. Pan et al.¹²¹ utilized molecular dynamics to study the adsorption and diffusion of oxygen atoms on Zr crystal surfaces. The results demonstrated that grain boundary orientation has a greater impact on oxygen adsorption rates than Frenkel defects, and oxygen diffusion is faster in defective crystals. Similarly, molecular dynamics simulations can also investigate the influence of substrate characteristics on oxidation behaviour. The basal plane (0001) of Zr exhibits greater resistance to oxygen diffusion compared to (10$\bar{1}$0) and (11$\bar{2}$0), resulting in superior oxidation resistance¹²². As the investigated scale expands, simulation methods for studying Zircaloys oxidation also include phase-field modelling¹²³. Combining experimental and computational approaches provides a more comprehensive understanding of the atomistic mechanisms of oxidation in Zr alloys, enabling the design of alloys with enhanced oxidation resistance through atomic-level control of adsorption and diffusion processes.

Outlook

This paper reviews the oxidation behaviour and mechanisms of Zircaloys, as well as factors influencing the phase transformation of zirconia films. It elaborates on the effects of alloy element, microstructure, and environmental conditions on oxidation kinetics, including atomic-scale simulations of oxygen adsorption and diffusion during Zr oxidation using molecular dynamics and first-principles calculations. Although existing studies have catalogued the oxidation kinetics of Zircaloys, there is a lack of microstructure-informed predictive models for oxidation kinetics. Current empirical formulas (e.g., parabolic rate law) exhibit significant limitations under high-temperature, high-burnup conditions, failing to guide the compositional design and process optimization of advanced Zircaloy. In addition, uncertainties persist regarding the phase evolution and defect formation mechanisms in oxide films, particularly the microstructural interplay between surface oxides and the underlying alloy substrate. In real-service environments, Zircaloys not only undergo oxidation in high-temperature, high-pressure water but also face mechanical stresses from fuel cladding swelling, neutron irradiation-induced crystallographic damage, and chemical interactions with boron/lithium ions in the coolant. Understanding how these factors synergistically regulate oxide film growth rates, phase composition, and defect density remains a critical challenge. Therefore, uncovering the microscopic mechanisms linking oxide phase transformations, defect evolution, and hydrogen permeation pathways-as well as clarifying the roles of oxygen vacancies, grain boundaries, and cracks in modulating oxidation kinetics and hydrogen embrittlement-will provide theoretical foundations and technical frameworks for developing next-generation accident-tolerant Zr alloys with enhanced oxidation resistance and reduced hydrogen pickup.