A detailed characterization of nanostructured thin zirconium oxide films formed during aqueous corrosion of a nuclear-grade zirconium alloy (Zircaloy-4) has been carried out by means of two novel, ...ultra-high-spatial-resolution grain mapping techniques, namely automated crystal orientation mapping in the transmission electron microscope (TEM) and transmission electron backscatter diffraction (t-EBSD). While the former provided excellent spatial resolution with the ability to identify tetragonal ZrO2 grains as small as 5nm, the superior angular resolution and unambiguous indexing with t-EBSD enabled verification of the TEM observations. Both techniques revealed that in a stress-free condition (TEM foil prepared by focused ion beam milling), the oxide consists mainly of well-oriented columnar monoclinic grains with a high fraction of transformation twin boundaries, which indicates that the transformation from tetragonal to monoclinic ZrO2 is a continuous process, and that a significant fraction of the columnar grains transformed from stress-stabilized tetragonal grains with (001) planes parallel to the metal-oxide interface. The TEM analysis also revealed a small fraction of size-stabilized, equiaxed tetragonal grains throughout the oxide. Those grains were found to show significant misalignment from the expected (001) growth direction, which explains the limited growth of those grains. The observations are discussed in the context of providing new insights into corrosion mechanisms of zirconium alloys, which is of particular importance for improving service life of fuel assemblies used in water-cooled reactors.
A study into the effects of irradiation temperature on the damage structures that form during proton-irradiation has been carried out on two commercial Zr alloys in order to develop a more ...mechanistic understanding of the effect of niobium on dislocation loop evolution. The two Zr alloys (Zircaloy-2 and Low-Sn ZIRLO™) were proton irradiated to a damage level of ∼2 dpa at 280 °C, 350 °C and 450 °C. Detailed dislocation analysis was carried out using on-axis bright-field scanning transmission electron microscopy combined with spectral imaging and synchrotron x-ray line profile analysis. The analysis revealed a significant difference in the effect of irradiation temperature on loop size between the two alloys. In the case of the Nb-free Zr-alloy (Zircaloy-2), an increase in irradiation temperature results in a marked increase in a-loop diameter, by a factor of ∼7.5 from 280 to 450 °C, and a stark decrease in the dislocation line density. In contrast, the Nb-containing Zr-alloy (Low-Sn ZIRLO™) showed very little variation of loop size and line density over the same radiation temperature range. The STEM-based spectral imaging revealed irradiation-induced nano-clustering found throughout the matrix in Low-Sn ZIRLO™, which is not present in the case of Zircaloy-2. Therefore, it is proposed that Nb plays a crucial role in the evolution of dislocation loops in Zr through the formation of irradiation precipitation throughout the matrix.
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Corrosion is a key limiting factor in the degradation of zirconium alloys in light water reactors. Developing a mechanistic understanding of the corrosion process offers a route towards improving ...safety and efficiency as demand increases for higher burn-up of fuel. Oxides formed on zirconium alloys are composed of both monoclinic and meta-stable tetragonal phases, and are subject to a number of potential mechanical degradation mechanisms. The work presented investigates the link between the tetragonal to monoclinic oxide phase transformation and degradation of the protective character of the oxide layer. To achieve this, Abaqus finite element analysis of the oxide phase transformation has been carried out. Study of the change in transformation strain energy shows how relaxation of oxidation induced stress and fast fracture at the metal–oxide interface could destabilise the tetragonal phase. Central to this is the identification of the transformation variant most likely to form, and understanding why twinning of the transformed grain is likely to occur. Development of transformation strain tensors and analysis of the strain components allows some separation of dilatation and shear effects. Maximum principal stress is used as an indication of fracture in the surrounding oxide layer. Study of the stress distributions shows the way oxide fracture is likely to occur and the differing effects of dilatation and shape change. Comparison with literature provides qualitative validation of the finite element simulations.
Detailed analysis was carried out on proton and a neutron irradiated Nb-containing Zr-alloy to study the evolution of dislocation loop size and densities as well as the formation and evolution of ...irradiation-induced precipitation/clustering. The results obtained here have been contrasted against previously published work on a Nb-free Zr-alloy 1, 2 to investigate the mechanistic reason for the improved resistance to irradiation-induced growth of Nb-containing Zr alloys. The combined use of bright field scanning transmission electron microscopy, ultra-high-resolution energy dispersive spectroscopy and atom probe tomography analysis provides evidence of evenly distributed radiation-induced Nb clusters that have formed during the early stage of proton irradiation and Fe-rich nano-rods near Fe-containing second phase particles. The former seems to have a profound effect on loop and subsequent loop formation, keeping loop size small but number density high while loops seem to initially form at similar dose levels compared to a Nb-free alloy but loop line density does not increase during further irradiation. It is hypothesized that the formation of the Nb nano-precipitates/clusters significantly hinders mobility and growth of loops, resulting in a small size, high number density and limited ability of loops to arrange along basal traces compared to Nb-free Zr-alloys. It is suggested that it is the limited loop arrangement that slows down loop formation and the root cause for the high resistance of Nb-containing Zr-alloys to irradiation-induced growth.
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The desire to improve the corrosion resistance of Zr cladding material for high burn-up has resulted in a general trend among fuel manufacturers to develop alloys with reduced levels of Sn. While ...commonly accepted, the reason for the improved corrosion performance observed for low-tin zirconium alloys in high-temperature aqueous environments remains unclear. High-energy synchrotron X-ray diffraction was used to characterize the oxides formed by autoclave exposure on Zr–Sn–Nb alloys with tin concentration ranging from 0.01 to 0.92wt.%. The alloys studied included the commercial alloy ZIRLO® (ZIRLO® is a registered trademark of Westinghouse Electric Company LLC in the USA and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited.) and two variants of ZIRLO with significantly lower tin levels, referred to here as A-0.6Sn and A-0.0Sn. The nature of the oxide grown on tube samples from each alloy was investigated via cross-sectional scanning electron microscopy. Atom probe analysis of ZIRLO demonstrated that the tin present in the alloy passes into the oxide as it forms, with no significant difference in the Sn/Zr ratio between the two. Synchrotron X-ray diffraction measurements on the oxides formed on each alloy revealed that the monoclinic and tetragonal oxide phases display highly compressive in-plane residual stresses with the magnitudes dependent on the phase and alloy. The amount of tetragonal phase present and, more importantly, the level of tetragonal-to-monoclinic phase transformation both decrease with decreasing tin levels, suggesting that tin is a tetragonal oxide phase stabilizing element. It is proposed that in Zr–Nb–Sn alloys with low Sn, the tetragonal phase is mainly stabilized by very small grain size and therefore remains stable throughout the corrosion process. In contrast, alloys with higher tin levels can in addition grow larger, stress stabilized, tetragonal grains that become unstable as the corrosion front continues to grow further inwards and stresses in the existing oxide relax.
Advancements in transmission electron microscopy allow us to draw correlations between evolving matrix chemistry environments and the resulting dislocation structures that form. Such phenomena are ...essential in predicting the lifetime of neutron reactor components, but are not well understood at the fundamental level. We investigate the effect of nano-scale matrix chemical evolution in Zircaloy-2 on dislocation formation after emulating commercial reactor irradiation conditions on a proton beamline. Similarity in the dislocation type, morphology, density and evolution between the different irradiation types establishes proton irradiation in this regard. For the first time, we observe chemical segregation of Fe, Ni and Cr to a-loop positions in basal traces and the segregation of Sn in alternate rows, anticorrelated to the positions of the light transition elements. The resulting layered structure with a periodicity of ∼50 nm creates an even greater anisotropy than that usually associated with HCP materials. Concurrent analysis of chemical effects and dislocation spatial relationships provides evidence that may explain the delayed onset of c-loop nucleation and accelerated dimensional instability regimes in its dependence on the alignment of a-loops parallel to the trace of the basal plane. This demonstrates the applicability of chemical-structural correlations towards key research questions regarding deformation behaviour.
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Pressurised and boiling water reactors contain zirconium alloys, which are used as nuclear fuel cladding. Oxidation of these alloys, and the associated hydrogen pick-up, is a limiting factor in the ...lifetime of the fuel. To extend the burn-up of nuclear fuel requires control of the oxidation, and therefore development of a mechanistic understanding of the cladding corrosion process. Synchrotron X-ray diffraction (S-XRD) has been used to analyse oxide layers formed during in-situ air oxidation of Zircaloy-4 and ZIRLO™. Analysis shows that as the oxide thickness increases over time there is a relaxation of the stresses present in both the monoclinic and meta-stable tetragonal phases, and a reduction in the tetragonal phase fraction. To better understand the mechanisms behind stress relaxation in the oxide layer, finite element analysis has been used to simulate mechanical aspects of the oxidation process. This simulation was first developed based on stress relaxation in oxides formed in autoclave, and analysed ex-situ using S-XRD. Relaxation mechanisms include creep and hydrogen-induced lattice strain in the metal substrate and creep in the oxide layer. Subsequently the finite element analysis has been extended to stress relaxation observed by in-situ S-XRD oxidation experiments. Finite element analysis indicates that the impact of creep in the oxide is negligible, and the impact of both creep and hydrogen-induced lattice strain in the metal substrate metal is small. The implication is that stress relaxation must result from another source such as the development of roughness at the metal–oxide interface, or fracture in the oxide layer.
Dislocation structures in neutron irradiated Zircaloy-2 fuel cladding and channel material have been characterized by means of high-resolution synchrotron x-ray diffraction combined with whole peak ...profile analysis and by transmission electron microscopy (TEM). The samples available for this characterization were taken from high burnup fuel assemblies and offer insight into the evolution of the dislocation structure after the formation of dislocation loops containing a c component. Absolute dislocation density values are about 4–15 times higher for the whole peak profile compared to TEM analysis. Most interestingly, the diffraction analysis suggests that the total dislocation density, as well as the a loop density, increases with fluence for the cladding material type. This trend is also inferred from a Williamson-Hall representation but contradicts the TEM observations. The c loop density evolution is more complicated and doesn't display any particular trend. In addition, the diffraction analysis highlights the presence of well-developed shoulders adjacent to the basal reflections and noticeable peak asymmetry particularly for the channel samples that experienced slightly lower operation temperatures than the clad. The findings are discussed in respect of the perceived irradiation induced growth mechanisms in Zr alloys.
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During corrosion of zirconium alloys a highly textured oxide is formed, the degree of this preferred orientation has previously been shown to be an important factor in determining the corrosion ...behaviour of these alloys. Two distinct experiments were designed in order to investigate the origin of this oxide texture development on two commercial alloys. Firstly, sheet samples of Zircaloy-4 were oxidised between 500 and 800 °C in air. The resulting monoclinic oxide texture strength was observed to decrease with increasing oxidation temperature. In a second experiment, orthogonal faces of Low Tin ZIRLO™1 were oxidised in 360 °C water, providing different substrate textures but identical microstructures. The substrate texture was observed to have a negligible effect on the corrosion performance whilst the major orientation of both oxide phases was found to be independent of substrate orientation. It is concluded that the main driving force for oxide texture development in single-phase zirconium alloys is the compressive stress caused by the ZrZrO2 transformation.
•Substrate orientation does not significantly affect oxide texture development.•Corrosion performance is independent of substrate texture.•Monoclinic oxide texture strength decreases with increasing oxidation temperature.•The main driving force for texture development is the oxidation-induced stress.
As a nuclear fuel cladding material, zirconium alloys act as a barrier between the fuel and pressurised steam or lithiated water environment. Controlling degradation mechanisms such as oxidation is ...essential to extending the in-service lifetime of the fuel. At temperatures of ∼360°C zirconium alloys are known to exhibit cyclical, approximately cubic corrosion kinetics. With acceleration in the oxidation kinetics occurring every ∼2μm of oxide growth, and being associated with the formation of a network of lateral cracks. Finite element analysis has been used previously to explain the lateral crack formation by the development of localised out-of-plane tensile stresses at the metal–oxide interface. This work uses the Abaqus finite element code to assess critically current approaches to representing the oxidation of zirconium alloys, with relation to undulations at the metal–oxide interface and localised stress generation. This includes comparison of axisymmetric and 3D quartered modelling approaches, and investigates the effect of interface geometry and plasticity in the metal substrate. Particular focus is placed on the application of the anisotropic strain tensor used to represent the oxidation mechanism, which is typically applied with a fixed coordinate system. Assessment of the impact of the tensor showed that 99% of the localised tensile stresses originated from the out-of-plane component of the strain tensor, rather than the in-plane expansion as was previously thought. Discussion is given to the difficulties associated with this modelling approach and the requirements for future simulations of the oxidation of zirconium alloys.
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