Zaheen Shah

Hydrogen uptake during aqueous corrosion of cladding zirconium alloys accelerates corrosion rate and leads to hydride formation, embrittling the cladding and causing significant safety concerns for reactor operation. Progress in understanding this process has been limited by the difficulty of probing hydrogen in oxides. Using deuterium tracing combined with atom probe tomography and complementary high-resolution characterisation techniques, we developed an analytical method for mapping hydrogen at near atomic scale in nanocrystalline insulating materials, demonstrated on zirconium oxide, opening new avenues for investigating hydrogen interactions in functional ceramics and solid-state battery materials. We show that oxide grain boundaries act as fast in-diffusion pathways for hydroxyl species, whereas separate nanopores act as traps that slow this process, indicating that controlling oxide grain width and porosity can help limit hydrogen ingress.

Sample preparation is critical to materials characterisation, with conventional mechanical preparation of zirconium alloys introducing a deformed surface layer that can reduce EBSD indexing quality and result in an incomplete representation of the microstructure. This paper investigates electropolishing of Low Tin ZIRLO™, identifying conditions that repeatedly achieved greater than 90% indexing. The resulting surface finish was compared with mechanical polishing using KAM and XRD to quantify differences. We show that poor surface finish can affect indexing of certain grain orientations, introducing bias in the measured grain population. These findings highlight the benefits of improved sample preparation and reporting parameters.

A novel Zr-Nb-Sn-V-Fe alloy has been studied using electron microscopy with energy-dispersive X-ray spectroscopy, atom probe tomography (APT), and X-ray diffraction to examine the role of V additions to Zr alloys used in the nuclear industry. The focus of this investigation is on solute distribution and second phase particles (SPPs), both of which influence the in-reactor performance of Zr alloys. Zr-V spatial distribution maps obtained via APT suggest that V is an interstitial element in α-Zr with a very low solid solubility on the order of 80 atomic ppm. Therefore, the majority of V precipitates out of solid solution as SPPs. Two distinct SPP types were identified. A body-centred cubic β-Nb phase containing small amounts of V (∼6 at.%) was observed to have a Nb- and V-enriched region surrounding the SPP approximately 4 nm from the interface. It is hypothesised that localised strains in the matrix around the SPP restrict solute transport from the matrix to the growing SPP. The second SPP type identified had an average composition of 30Zr-42V-14Nb-13Fe (at.%) determined via APT, although the chemical composition was heterogeneous throughout individual SPPs. Based on the measured composition, a stoichiometric designation of Zr(V,Nb,Fe)2 is proposed. Crystallographic analysis showed structural heterogeneity within the SPPs, with evidence indicating the hexagonal C14 Laves phase. The observed chemical and structural variations suggest that this phase is metastable or undergoing a sluggish phase transformation during the final heat treatment. These findings help further the understanding of material microstructure and phase transformations in next generation Zr cladding alloys.

Ex-reactor ZIRLO and AXIOM X2 Zr alloy samples exposed to up to 5.2 years of in-reactor irradiation and corrosion were studied using atom probe tomography to understand the effect of irradiation on the corrosion of the cladding Zr alloys. The solute redistribution during irradiation depends strongly on the starting microstructure, where Fe and Nb form clusters aligned the basal plane of the recrystallised X2 alloy whereas Fe, Nb and Sn segregate on the columnar grain boundaries in the stress-relieved ZIRLO alloys. The distinct microstructure and the solute redistribution during irradiation can be related to the degraded irradiation swelling and in-reactor corrosion resistance of ZIRLO compared to X2. During in-reactor corrosion, the oxide front grows into an irradiated metal substrate that contains these chemical heterogeneities, but most rapidly become redistributed once incorporated into the oxide. Nb clusters, as an exception, still persist ~1 µm into the oxide, but are gradually dissolved releasing Nb to the surrounding oxide, and are finally oxidised to (Zr,Nb)O2. The effect on subsequent corrosion kinetics of changing the Nb content in the Nb clusters and in the surrounding oxide is discussed in detail. These results are valuable towards understanding the coupled effects of irradiation and corrosion, and so developing systematic model of in-reactor corrosion.

In this study we have used a combination of atom probe tomography and (scanning) transmission electron microscopy to investigate solute redistribution under high fluence neutron irradiation in a Zr-Nb-Sn-Cu-Fe experimental fuel cladding alloy. The bulk samples were exposed to neutron irradiation in the BOR-60 reactor up to 33.5 dpa. By mapping solute distributions and quantifying solute compositions in carefully defined regions of irradiated materials, we show that during irradiation Fe and Cu are gradually depleted from second phase precipitates, grain boundaries and interfaces present in the unirradiated state, forming characteristic bands of enrichment at arrays of microstructural defects on the basal plane. In-solution Sn and Nb concentrations in the unirradiated material are depleted to form Sn bands that are spatially out-of-phase with Fe and Cu bands, and Nb clusters which agglomerate and coarsen with longer irradiation exposure. The compositions of both the original β-Nb second phase precipitates and the irradiation-induced Nb clusters converge on ~Zr-35%Nb after heavy irradiation. These results provide new insights into solute redistribution during irradiation. The significant microchemical variations observed at high doses can be used to refine mechanistic understanding of the coupled irradiation and corrosion processes of cladding under in-reactor conditions.

Samples of the AXIOM X2 Zr alloy exposed to neutron irradiation and corrosion in the Vogtle reactor for five years were studied using atom probe tomography. We have shown that the chemistry of the metal substrate below the oxide layer formed during aqueous corrosion has been substantially changed by the incorporation of Li from the coolant water, and, with the support of FISPACT-II inventory calculations, the formation of radioactive 94Nb and Mo isotopes as transmutation products from Zr and Nb. The findings highlight the importance of investigating the degradation mechanisms of Zr cladding alloys under real service conditions, and raise new questions on how these irradiation-driven chemical changes can influence both the in-service corrosion performance and the management of the radioactive waste at end of life.

The addition of copper (Cu) in zirconium (Zr) alloys remains relatively uncommon, with existing research primarily focused on its influence on enhancing corrosion resistance. This work examines the microstructural evolution of the Zr-1Nb-0.3Sn-0.2Cu-0.05Fe (W13) alloy during neutron irradiation. Transmission electron microscopy and atom probe tomography were employed to characterise the microstructure of the alloy in both the unirradiated condition and after neutron exposure at doses reaching up to 33.5 displacements per atom. In the unirradiated state, secondary phase particles (SPPs) of beta-Nb, Zr(Nb,Fe)2 and Zr2(Cu, Fe) SPPs were identified. Neutron irradiation resulted in Cu depletion from the Zr2(Cu, Fe) SPPs, with the released Cu forming solute bands co-segregated with iron (Fe) and anti-segregated with tin (Sn). The beta-Nb SPPs exhibited a decrease in niobium (Nb) content and an increase in size, attributed to Zr incorporation. During irradiation, Nb-rich SPPs formed in rows along basal planes, with some containing Fe. Mechanisms were proposed to explain SPP formation, solute depletion from SPPs and the formation of solute bands specific to the W13 alloy.