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Over the last few years, TEM has become a powerful technique to study cracks and specially crack tips. However, the number of publications including TEM results has not grown as it was expected. The main reason for this might be difficulties in the sample preparation. In this work we present a novel FIB sample preparation technique which has proved to be an ideal tool for preparing cross sectional samples containing crack tips. The morphology of intergranular stress corrosion cracking (IGSCC) has been investigated in Alloy 600 subjected to constant load and slow strain rate tests in simulated primary circuit pressurized water reactor conditions. Cracks were observed to nucleate at high-angle grain boundaries and propagate to depths of a few tens of micrometer along such boundaries, still in the initiation stage. Electron diffraction, energy dispersive x-ray (EDX) and electron energy loss spectroscopy (EELS) have been used to identify the different corrosion products and precipitates. Elemental mapping was employed to reveal changes in composition in the crack tip area. Major observations at cracks and grain boundaries include: the presence of different oxides in different locations, differences in grain boundary oxides and open crack/free surface oxides. These observations suggest that IGSCC involves oxygen diffusion through a porous oxide region along grain boundaries to the bare metal. This is a novel concept that offers an alternative to previous mechanisms proposed in the literature e.g. H embrittlement, slip-dissolution, etc., for which no supporting evidence has been found.

Intergranular Stress Corrosion Cracking (IGSCC) has been one of the most important challenges to nuclear power plant operation worldwide. The degradation of alloy 600 (Ni-based alloy) and its susceptibility to Stress Corrosion Cracking (SCC) when exposed to primary water environment in pressurized water reactors (PWRs) has been a topic of interest for many decades among the SCC community. However, the understanding or prediction of SCC mechanisms remains questionable. To address this, an IGSCC predictive model called 'Local Model' has been developed in the framework of EDF's CSI project. The main focus of this research is to investigate IGSCC in alloy 600 and to experimentally validate a predictive model. Firstly, it is important to know whether stress corrosion cracking occurs in these materials, in the way presented in the existing experiments and models in the literature. These have shown a very large scattering for the needed for the crack initiation and crack propagation, indicating significant model uncertainties. In this thesis, an experimental and modelling program has been developed for investigating SCC in alloy 600 for conditions simulating PWR conditions under constant load. The goal of the experiments was to find a reliable and quantitative determination of crack initiation and crack propagation for these conditions. For the modelling aspect, the aim was to determine the critical parameters of SCC by considering the phenomena in the local scale of grain boundaries and crack tips. The model requirements include the ability to predict the crack initiation and to model crack growth rate as a function of macroscopic and measurable parameters. This has been validated by experiment to ensure that the assumptions of the model are realistic enough to make sure of reliability for reactor safety applications. By doing this, the model allows for a better understanding of the underlying mechanisms of SCC. As such, it focuses on the phenomenological approach rather than the empirical approaches in other models.

Polymer electrolyte membrane fuel cells (PEMFCs) are promising energy-conversion devices due to their high efficiency and near-zero carbon emission footprint. PEMFCs still require a high mass of precious metal Pt as an electrocatalyst for oxygen reduction reaction (ORR) to ensure sufficient power output efficiency, but the sources and cost of Pt greatly limit its commercial application in fuel cells technology. The development of high-active and low-cost ORR electrocatalysts is therefore of much significance for the commercialization of fuel cells. Compared to Pt, bimetallic (Pt-Co) nanoparticles exhibit superior catalytic performance with less mass of precious metal, but the intrinsic mechanism of this catalytic enhancement has not been fully explained. Strain and composition are two fundamental effects that are highly associated with the catalytic activity of the Pt-based nanoparticles. Thus, being able to measure local composition and strain at an atomic scale within the individual nanoparticle will further our understanding of the operation of heterogeneous catalysts. In this thesis, the strain analysis and composition quantification are precisely and simultaneously performed on the individual nanoparticle at an atomic scale for four Pt-Co nanoparticle ensembles, to relate composition and size to strain. We find that no compressive strain on the Pt-rich outer shell is imposed by the alloy core. Dilation strain is found to be distributed throughout the individual nanoparticles, with the radial dilation strain being much larger than the circumferential dilation strain at the surface. Remarkably, all the Pt-Co stoichiometries studied have a relatively invariant surface lattice parameter, which is smaller than that for pure Pt but larger than that predicted using the measured local composition, whether or not an acid de-alloying treatment is applied. This work makes it possible to design the optimum ORR catalysts by tuning strain, composition, and structure of the Pt-based bimetallic nanoparticles.

Austenitic alloys are commonly used in the nuclear industry due to their high corrosion resistance and excellent mechanical properties. De- spite boasting an impressive service history, it is possible for critical structural components made from these alloys to undergo stress cor- rosion cracking (SCC) under pressurised water reactors (PWR) condi- tions. As the primary method of material degradation, SCC has been studied extensively for several decades, and many factors have been shown to affect SCC susceptibility, these include cold-work, tempera- ture and water chemistry. To develop a better mechanistic understand of SCC further research is required. This body of research evaluates the effect of hydrogen on mechanical deformation in SCC. Localised and high-resolution testing has been employed successfully to characterise key processes in SCC crack propagation. For this reason, nanoindenta- tion, transmission Kikuchi diffraction (TKD), NanoSIMS and Thermal desorption spectrscopy (TDS) have been used to characterise material deformation on a length-scale relevant for SCC. The main objective of this thesis was to correlate the SCC crack growth rate's (CGR) dependency on nickel content, first observed by Coriou, with both the localised mechanical deformation and hydrogen up-take at PWR temperatures. To satisfy this objective three alloys have been systematically tested in every chapter, these include SS316L (12%- Ni), A800 (32%-Ni) and A600 (72%-Ni) as they represent commercially available alloys with low, intermediate and high nickel contents. Room temperature micromechanical testing has been used to compare the mechanical properties of austenitic alloys on a length-scale that is relevant to SCC. The mechanical properties of all austenitic alloys are highly comparable at room temperature, and show no deviation with changing nickel content. High-magnification cross-sectional analysis shows that the deformation mechanisms in A800 differ from SS316L and A600 at room temperature. A800 does not undergo Σ3 defor- mation twinning, but instead deforms by dislocation glide. This is unexpected as the stacking fault energy of A800 is an intermediate of the other alloys. Deformation is a main SCC mechanism, and the ob- served alloy deformation correlates with Arioka's reported SCC CGR in PWRs, i.e the slowest CGR undergoes no twinning. As the me- chanical properties of alloys with a 25-40% nickel content do not differ significantly from other austenitic alloys, it is thought that their corro- sion resistance is weighted towards diffusion-based SCC mechanisms. Such an explanation is consistent with findings at hotter temperatures, where diffusion-based mechanisms are more active. High temperature micromechanical testing supports these observations, and can be used to better understand the individual mechanism weights depending on alloy composition and temperature. NanoSIMS and TDS were used to quantitatively and qualitatively ex- plore the hydrogen trapping in austenitic alloys exposed to pure heavy water at PWR temperatures. Hydrogen is found trapped by various microstructural features, such as carbides, grain boundaries and ox- ides by NanoSIMS. The total hydrogen content of SS316L and A800 is five-times greater than that of A600, which correlates with the en- thalpy of diffusion in the alloys. These results show an inverse trend with the SCC failure reported by Coriou, and a weaker inverse correla- tion to Arioka's results. The weaker correlation with Arioka's results is due to a change in the corrosive medium from pure water, used in this study and by Coriou, to simulated primary PWR water. It is suggested that the inverse relationship could be due to the elastic shielding pro- vided by the hydrogen enhancing dislocation motion, and preventing the stress intensity factor reaching a critical values that will initiate crack propagation.

This thesis, presents results from the Joint European Torus (JET), beryllium ITER-like wall. Isotopic hydrogen retention is investigated at three microstructurally unique sites: erosion zones, co-deposition zones, and melted material, found across the reactor at the inner and outer wall limiters and dump plate tiles. The techniques applied to these novel inservice materials were: Scanning Electron Microscopy with Energy Dispersive X-Ray (SEM-EDX), Raman Spectroscopy; Focused Ion Beam (FIB) 3D milling; Scanning Transmission Electron Microscopy (STEM) with Electron Energy Loss Spectroscopy (EELS) and Energy Dispersive X-Ray (EDX); Atom Probe Tomography and Thermal Desorption Spectroscopy (TDS). Clustering of deuterium (D) with beryllium deuteride (BeD2) was found at the outer eroded material, whilst not at the inner. BeD2 was present at the core of D clusters up to 8 at%. Co-deposition at the outer limiter tile was governed by beryllium oxide (BeO), whilst at the inner it was governed by Be with D. EUROFusion samples produced from High Power Impulse Magnetron Sputtering (HiPIMS) returned similar EELS results to JET samples; presenting a reasonable model. Melted material produced highly oxidised zones devoid of metallic impurities from marker layers in spatially optimum sites, surrounded by grains produced via high temperature material mixing. Several types of beryllium deuteroxide (BeOD) with varying stoichiometries were presented Retention is lowest at the inner wall eroded zone, increasing at the outer wall eroded zone, and increasing further between inner to outer wall codeposits. Melted material, remains an outlier. Overall total retention, in melted materials was similar to outer wall co-deposit, however composed of over 77% H. Material mixing was suggested the cause due to beryllium/nickel alloying seen in melted materials.

Zirconium alloys have been successfully used as fuel cladding and structural materials in fission reactors for decades, but the industry’s desire to employ high burnup fuels requires an improvement of resistance to in-reactor corrosion. It is therefore essential to understand the mechanisms of radiation-enhanced corrosion of zirconium alloys, and especially why Zr-Nb alloys show improved properties compared to Zircaloys. In order to achieve this goal, the MUZIC-3 (Mechanistic Understanding of Zirconium Corrosion) program was established; a wide collaboration between universities and industrial partners. As part of this program, this project used advanced electron microscopy techniques to investigate in-reactor corroded samples from the Canada Nuclear Laboratory and autoclave corroded samples from Westinghouse. The characterisation work used the (S)TEM as a primary tool, together with the analytical tools EDX and EELS. The in-reactor samples were Zr – 2.5 Nb samples corroded at 325 °C and 250 °C in the reactor core for 190 days and 2750 days. For comparison, samples corroded in the same primary loop but out of the reactor core, and hence without neutron radiation, have also been studied. In order to further study the mechanism of radiation damage in the metal matrix, in-situ heavy ion radiation in a TEM was carried out on the recrystallised Zr – 1.0Nb and Zr – 2.5Nb alloys provided by Westinghouse™, and proton irradiation was carried out at Dalton Cumbrian Facility (DCF) on as-received Zr – 2.5Nb alloys from CNL. All the TEM samples in this project were made by Focused Ion Beam (FIB) at either the University of Oxford or the Materials Research Facility (MRF) at the Culham Science Park. The first section of results presented in this thesis is the characterisation of the metal matrix of the CNL Zr – 2.5Nb alloy. The shape, orientation and number density of radiation-induced Nb precipitates in the α-Zr grains in samples corroded under neutron radiation at 325 °C was analysed. The direct observation of fully coherent Nb nano-precipitates in the α-Zr metal matrix is reported for the first time. A model describing the formation and evolution of the Nb precipitates under neutron irradiation is suggested. The impact on the evolution of the β-Zr phase from the in-reactor conditions is also presented, and it is shown that decomposition from β-Zr to β-Nb is encouraged by neutron irradiation and in-reactor heating. The second section of the results discusses the mechanisms of radiation damage to the β-Nb SPPs in the metal matrix, as studied by in-situ heavy ion radiation. It was found that due to the combined effect of radiation-induced mixing and local diffusion of solute Nb atoms, the β-Nb SPPs stay very stable under irradiation up to 40 dpa and show very limited Nb redistribution between the SPPs and α-Zr metal matrix. Reasons for the remarkable stability of these Nb-rich phases under radiation are discussed. The characterisation of proton irradiated Zr – 2.5Nb samples is also presented. The results help to understand the mechanisms of radiation-induced Nb precipitation under a variety of conditions. The third section presents a study of the corrosion of the decomposed β-Zr regions in the oxide. This is a systematic study of the corrosion behaviour, including morphology changes, oxidation rates of the Nb in the β regions, redistribution of the Nb, and the relation with the formation of other microscopic features such as micro cracks. It is shown that the longer these β regions stay in the metal under neutron radiation, the more decomposed from β-Zr to β-Nb phase they become, and the slower that these regions are oxidised. With all these factors combined together, the reasons for the improved resistance of the Zr-Nb alloys to in-reactor corrosion is discussed.

This thesis is part of the MUZIC-3 (Mechanistic Understanding of Zirconium Corrosion) project, with the overall goal to understand the in-reactor corrosion and hydrogen pick-up mechanisms of zirconium alloys. Zirconium alloys are commonly used for fuel cladding and support structures in light water reactors but suffer from aqueous corrosion in service that can limit the operating lifetime and the effective burnup of the fuel. Understanding the mechanisms of corrosion, especially under irradiation, is thus of great importance for the development of accurate, physically based lifetime prediction models for cladding materials, and providing information for the design of new corrosion-resistant alloys. Acceleration of the corrosion rate has been widely reported as a result of exposure to in-reactor conditions, but the mechanisms of this accelerating effect are still not well understood. It has been correlated with the specific in-reactor coolant chemistry, intense γ-radiation or irradiation-induced redistribution of alloying elements, but rather little attention has been paid to the irradiation-induced degradation of the corrosion protective oxide layers. In an effort to understand such mechanisms, I have used both in-situ radiation damage techniques and direct observations of ex-reactor materials to study radiation effects in zirconium oxides formed on a range of Zr alloys. The aqueous corrosion of zirconium follows the pathway: Zr→h-ZrO→ZrO2. The transformation from Zr to h-ZrO is found to follow a displacive mechanism, and the suboxide grains can show different morphologies, equiaxed, plate-like or sawtooth-like, depending on the underlying α-Zr grain orientation, although no specific orientation relationships between h-ZrO and m-ZrO2 or α-Zr and m-ZrO2 were identified. The monoclinic-ZrO2 oxides formed in an autoclave are observed to have a similar texture on different alloys, with (10̅4)m-ZrO2 parallel to the metal/oxide interface regardless of the underlying α-Zr or h-ZrO orientations. I report for the first time on the susceptibility to radiation damage of the suboxide phase which may influence the nucleation of new oxide grains and the transportation of oxidation species across the oxide/metal interface, and lead to enhanced corrosion rates. A monoclinic-to-cubic transformation of the bulk oxide is also observed by in-situ ion irradiation experiments, followed by irradiation-induced grain growth. The possibility of radiation-induced stabilisation of this cubic phase thus needs to be considered as a possible process that can occur at high burnups in reactors, and may further affect the corrosion rates. As a result of the in-reactor corrosion conditions, e.g. irradiation and water chemistries, the oxides formed in-flux are less well textured and with a more disrupted grain, which can contribute to enhanced corrosion rates in-flux. My results also indicate that oxide nano-porosity plays an important role in the transportation of oxidising species throughout the oxide layer. The density of nano-porosity in the oxide corrosion layers has been successfully quantified for the first time as function of both depth in the oxide and exposure time, and a clear correlation is observed between the measured increase in hydrogen pickup fraction and the characterised increase in interconnected porosity in the oxide. To complete the work in this thesis, a wide range of microstructural characterization techniques have been applied, like Focused Ion Beam, Transmission Electron Microscopy, Transmission Kikuchi Diffraction and high-resolution Secondary Ion Mass Spectrometry. The combination of these techniques has provided me with a wide range of information on microstructures and microchemistry, which is useful to the understanding of the degradation mechanisms in fuel cladding materials.

In the absence of a high-flux fusion-neutron irradiation source, the microstructural and mechanical changes expected within materials exposed to a nuclear-fusion environment must be replicated by fission-neutron and other surrogate-particle irradiations. This study uses transmission electron microscopy (TEM) to compare the microstructural defects produced in FeCr alloys during exposure to neutrons, protons, and self-ions. Alloys of Fe6Cr and Fe9Cr were irradiated using fission-neutrons, 2.0MeV Fe+ ions and 1.2MeV protons at similar temperatures (~300C) and similar doses (~2.0dpa). The neutron-irradiated alloys contained a population of interstitial dislocation loops with b=<111> (>70%) and b=<100>. The visible dislocation loops were on average ~5nm in size, and the density varied from 2±1 x1014cm-3 in the matrix to 1.2±0.3 x1017cm-3 close to helical dislocation lines. Dislocations loops were mostly clustered around sub-grain boundaries and helical-dislocations. Helical-dislocations formed from initially straight screw dislocations experiencing radial-climb in response to a vacancy-biased defect flux. Small chromium clusters were identified in the neutron-irradiated Fe6Cr, and chromium α'-phase precipitates were identified in the Fe9Cr. Self-ion irradiation produced mostly homogeneously distributed dislocation loops (6-7nm on average), but with a greater fraction of <100> loops (~40%) than was seen in the neutron-irradiated alloys. The self-ion irradiated Fe6Cr and Fe9Cr contained only vacancy-type loops, unlike the neutron or proton irradiated sample which contained only interstitial loops. Chromium remained in solution in both ion-irradiated samples. Proton-irradiated Fe9Cr contained dislocation loops close to helical-dislocation segments, similar to the neutron-irradiated sample. Chromium α'-phases were also identified. The proton-irradiated Fe6Cr contained much larger loops (~13nm on average) than the neutron or ion-irradiated alloys, and chromium was shown to have segregated on and around these loops. Both proton-irradiated alloys contained large voids (>4nm and up to 12nm) at a density greater than 1016cm-3. In the neutron and ion-irradiated alloys, voids were mostly <2nm.

Zirconium alloys are widely used as structural and fuel cladding alloys in current nuclear pressurised water reactors. The MUZIC-2 programme, which this project has been part of, is a collaboration between an international team of universities and industry in furthering the understanding of oxidation and hydrogen pickup of these alloys. This project has contributed to the programme by analysing a large number of oxidised zirconium samples by scanning transmission electron microscopy (STEM), scanning electron microscopy (SEM), 3D focused ion beam (FIB) sectioning and secondary ion mass spectrometry (NanoSIMS). By STEM observations, ZIRLO and Zircaloy-4 have been analysed at different stages of oxidation, comparing oxide thickness, crack content, porosity and second-phase particle content. Utilising state-of-the-art fast low-loss electron energy loss spectroscopy (EELS), the oxygen-saturated metal region has been mapped. The band of oxygen-saturated metal below the metal-oxide interface is found to be thinner for samples that have recently finished the fast stage of oxidation (early pre-transition and early post-transition) than samples in the slower stages of oxidation (late pre-transition). Core-loss EELS has been fitted using a low-loss convolution curve fitting approach and has been used to measure the energy position and intensity of the Zr L edges and white lines across the metal-oxide interface. As the Zr oxidises, the white lines are found to shift to higher energy loss, and the ratio between the L3 and L2 white lines increases. The metastable ZrO suboxide phase is found to follow this trend. Novel hydrogen and deuterium measurements made using NanoSIMS have been performed on deuterium-traced alloys in both cross-section and as three-dimensional depth profiles to understand the ingress of hydrogen through the oxide barrier. The pre-transition oxide is found to act as an adequate barrier to hydrogen ingress whilst the post-transition oxide is found to be porous and only protective close to the metal-oxide interface. A hydrogen uptake mechanism hypothesis is proposed. Energy-dispersive X-ray spectroscopy suggests that second phase particles may constitute a possible route for hydrogen ingress when close to the metal-oxide interface. Together, these observations contribute to improving the understanding of the corrosion of zirconium alloys.

Austenitic alloys have been extensively used in the nuclear industry as structural components due to their combination of excellent mechanical properties and high corrosion resistance. Although these alloys have a good service record, many can become susceptible to stress corrosion cracking (SCC) under pressurized water reactor (PWR) primary water conditions. After several decades of study, a considerable number of factors have been revealed that affect SCC susceptibility, such as material composition, pre-existent cold-work, temperature or water chemistry. Although the effects of these factors on SCC have been extensively studied, a better mechanistic understanding is still needed. High-resolution characterization techniques can reveal key processes involved in SCC crack propagation. For this reason, analytical transmission electron microscopy (ATEM) and transmission Kikuchi diffraction (TKD) have been used to characterize the crack tip region. Several SCC precursors have been identified: intergranular oxidation zone (IOZ) ahead of the crack tip, Cr-Fe depletion ahead of the IOZ, grain boundary migration along the crack and ahead of the crack tip, and localized deformation around the crack tip. The first objective of this thesis was to understand the underlying connections between these precursors and SCC propagation. Intergranular carbides and grain boundary migration were observed to be able to restrain SCC propagation. In addition, two mechanisms were confirmed to play important roles in the SCC propagation: one being a diffusion-based and the other one mechanical deformation-based. The second objective of this thesis was to obtain a better mechanistic understanding of the temperature dependence of crack growth rate (CGR) in 316SS and Alloy 600 in the temperature range of 320-360ºC. It was found that the two proposed mechanisms seem to have different weights on each alloy at "low" and "high" temperatures. The third objective was to explain the effect of Ni content on SCC CGR, also known as the "Coriou effect". Four alloys with different content of Ni, including 316SS, Alloy 16Ni, Alloy 60Ni, and Alloy 600 were characterized. The results showed that although the elemental diffusion increased with the increase of Ni content in the Fe-based alloys (316SS and Alloy 16Ni), the mechanical deformation-based mechanism appeared to play a dominant role, leading to the CGR decrease with the increase of Ni content in the Fe-based alloys. Both the diffusion-based mechanism and the mechanical deformation-based mechanism increased the CGR in Ni-based alloys (Alloy 60Ni and Alloy 600) with the increase of Ni content. The excellent SCC resistance of the alloys in the range of 20-45%wt. Ni was mainly contributed by the low extent of localized deformation around the crack tip, although the elemental diffusion in these alloys was not very low.

The thesis aims to develop methods to explore how catalyst structure and composition affect catalytic performance of nanoparticles used in Proton Exchange Membrane Fuel Cell (PEMFC)applications. Using Scanning Transmission Electron Microscopy (STEM), the structure and composition of Pt and Pt-Co catalysts are explored with the intent of designing better catalysts. Catalysts such as pure Pt and alloyed Pt-Co ensembles are hugely popular in the PEMFC industry due to their catalytic activity and stability. Out of the two ensembles, the bimetallic Pt-Co catalysts exhibit better performance and are cheaper to manufacture. The addition of a secondary element reduces the overall Pt loading making the catalyst cheaper. Introducing a secondary alloy also results in enhanced catalytic activity for reasons not well explained. To study the effects of enhanced activity, a thorough characterisation at the atomic scale is required. This means studying a catalyst's size, shape, strain and composition simultaneously, at high throughput, and linking this to catalytic activity. Full characterisation of this detail is challenging, but not impossible. Using STEM, it is possible to obtain catalyst size, shape, strain and composition simultaneously. However due to the nature of the technique is it challenging to characterise industrially relevant materials at high throughput. This thesis develops upon current methods for characterising nanoparticles in STEM to obtain nanoparticle size, shape, strain and composition at high throughput. In STEM, the size, shape and strain information is obtained from the Annular Dark Field (ADF) signal, whereas the composition information is obtained from Energy Dispersive X-ray Spectroscopy (EDS) and Electron Energy Loss Spectroscopy (EELS). One of the challenges is combining the information from ADF, EDS and EELS signals into one quantifiable unit and building an analysis framework for catalyst nanoparticles. This can be achieved by converting the measured STEM signal into scattering cross-sections. The scattering cross-section describes the effective area corresponding to the probability of scattering from a sample. With careful microscope calibration, the scattering cross-section can be converted to number of atoms. Atom counts from the STEM signals provide the structure and composition information from a nanoparticle. The thesis will explore the minimum requirements needed to develop new automated methods to characterise structure and composition of nanoparticles using scattering cross-sections at high throughput. In addition, limitations and capabilities of current hardware and software are explored. The three-dimensional models obtained from the quantitative analysis can be used as inputs for Density Functional Theory (DFT) simulations to understand molecule adsorption mechanics better. Combining high throughput technique development, STEM quantification and high scale DFT simulation together unlocks a detailed insight in understanding catalyst activity. Such detailed description of a system is the stepping stone towards reducing cost and time for better catalyst design.

Copper (Cu) has high thermal conductivity and is thus ideal for high heat flux, thermal heat sink applications in fusion power applications. Divertor designs for future fusion power plants will expose Cu alloys to extreme thermal (>10 MW m-2) and neutron fluxes (200 dpa) that destabilise the microstructure of Cu. To improve stability, strength and creep resistance, alloying additions are used commercially, but these compromise thermal conductivity. Dispersed oxide particles may offer the opportunity for improved mechanical stability and creep resistance even at very low volume fractions (<1%) while avoiding large reductions in thermal conductivity. However there are few studies on the processing-performance of oxide dispersion strengthened Cu alloys. In this thesis, novel oxide dispersion strengthened Cu alloys were prepared by room temperature mechanical alloying of Y2O3 and the mechanism of dispersion investigated. A small fraction of Y2O3, up to 1%, was shown to disperse effectively during mechanical alloying at room temperature in Cu. Both severe fragmentation and some local disassociation of the Y and O occurred, allowing for re-precipitation of fine nanoparticles <10 nm during consolidation and exposure to elevated temperatures. A model alloy of Cu-2 wt.% Y2O3 alloy had a mean oxide particle diameter of 7.1 ± 6.0 nm and a mean particle number density of 8.24 x 1022 m-3 following consolidation. Ternary Ti additions were investigated for nanoparticle refinement and best design alloy with a composition of Cu-0.25Y2O3-0.25Ti was produced that had a mean nanoparticle diameter of 3.2 ± 1.5 nm and a mean particle number density of 2.36 x 1023 m-3 , which after thermal ageing for 545 h at 350 °C, was largely unchanged at 3.8 ± 1.7 nm, and 1.74 x 1023 m-3 respectively. Comparing favourably with commercial Al2O3 dispersion strengthened Cu, the alloy had a narrower particle size distribution and a higher particle number density. The fine dispersed oxide nanoparticles gave good grain boundary pinning, retaining an ultrafine mean grain size of 220 nm after thermal ageing. Thermal conductivity of the Ti-containing alloy was 332 ± 16W m-1 K-1 , and the addition of Ti increased the thermal conductivity with increasing temperature. The creep resistance was evaluated by small punch testing and in-house produced alloys outperformed commercial alloys at 350 °C. The work in this thesis indicates that mechanically alloyed Cu-Y2O3 or Cu-Y2O3-Ti alloys, with further development and evaluation, have potential as thermal heat-sink materials for fusion divertor application.

The oxidation of zirconium alloys under aqueous conditions has been studied for more than 50 years in relation to its role as a fuel cladding element in water-cooled nuclear reactors. The most well-known and interesting phenomenon associated with the corrosion of zirconium alloys is the cyclic oxidation rate. Oxidation initially proceeds rapidly, forming an oxide film. As this film thickens, the rate of oxidation slows before transitioning, and begins oxidising at the higher rate again. This effect has been well studied, although the cause of rate transition is not understood. Using samples oxidised in conditions very closely matching those of a nuclear reactor, although without radiation, several techniques were used to characterise the oxide before and after the transition in oxidation rate. Comparison of these results allows some insight into the mechanism operating. Doping samples with isotopic tracers during different points in the oxidation cycle and subsequent analysis with NanoSIMS gave insight into the porosity and protectiveness of the oxide film before and after transition. Before transition, the entire oxide film acts as a barrier, rather than a small sublayer. After transition, the whole layer of previously formed oxide became porous and the oxidising media was admitted directly to the metal/oxide interface. Examining samples of differing oxidation times showed each subsequent transition was a close repeat of the first. Oxide thickness, rather than time under corrosion conditions, was identified as the best predictor of whether a sample would undergo transition. Oxide porosity (and thus rate transition) was demonstrated to be a local process on scales as small &LT; 1μm. Residual stress in the oxide and underlying metal substrate have been studied for zirconium alloys for some years. Here, a new technique is applied that largely confirms previous measurements made by XRD. Testing pre- and post-transition oxides revealed a large drop in the compressive stress after transition, showing the post-transition oxide is both porous and contains much lower compressive stresses.

This thesis was part of the MUZIC-2 collaboration and describes the effect of different water chemistries on the corrosion and hydrogen pick-up in Zircaloy-4. Samples of Zircaloy-4 were oxidised in an autoclave containing pure water at 360°C (pH360 = 6.15) and at 350°C in an elevated pH (pH350 = 8.82, with 50% deuterated water) compared to commercial reactors. They were examined by scanning transmission electron microscopy (STEM), Thermal Desorption Spectroscopy (TDS) and Secondary Ion Mass Spectroscopy (SIMS). There was a change in the oxidation rate for the samples exposed to different environments, in part from the difference in the oxidation temperature and in part due to the pH of the oxidising environment, although it is the elevated pH which has a lower oxidation rate in the pre-transition regime. The oxide-metal (OM) interface was characterised along with second phase particles (SPPs) present in the metal and the oxide. It was found that the region of oxygen-saturated zirconium (OSZ) and ZrO follow a cyclical growth pattern which has the same periodicity as the advance of the OM interface. The OSZ gradually thickens as the oxidation rate decreases in the late pre-transition period, before disappearing after transition and then growing again in the second cycle of oxidation. The thickness of the layers formed appears to be independent of the pH used in the experiments. The oxidation of Fe and Cr and the Fe/Cr ratio in SPPs were examined. The SPPs were oxidised after the surrounding Zr oxidised. The Fe/Cr content in the middle of the SPPs decreased quickly as the Fe was found to migrate towards cracks formed at the edge of SPPs. Cr migrated much more slowly towards the cracks than Fe and oxidised before Fe. The hydrogen/deuterium content of the samples was measured with TDS. Two characteristic desorption peaks for hydrogen have been found at ~350°C and ~650°C. The latter occurs when the difference in free energy between hydrogen in the metal and in the gas phase becomes positive. There is a delay to the desorption when an oxide layer is present, until the temperature reaches the point at which it is dissolved into the metal, allowing the hydrogen to desorb. The hydrogen pick-up during the first 1.5 μm of oxide growth was similar for both environments. For thicker oxides the samples exposed to pure water showed four times the amount of hydrogen pick-up compared to the high pH samples. The modelling of hydrogen desorption from Zr was carried out using the Tritium Migration and Analysis Program (TMAP). It was possible to replicate the low temperature peak with the literature values for the binding energy of hydrogen with Zr and the hydrogen diffusion activation energy (ED), but for the high temperature peak the required ED was higher than the literature value. An attempt at modelling both peaks simultaneously is shown.

Zr cladding alloys have been used for many years as the first safety barrier layer of a nuclear reactor. However, the recent Fukushima accidents and industrial demands to increase the burnup of fuels have led to increasing interest in a detailed mechanistic understanding of aqueous corrosion and hydrogen pickup and the performance at high temperatures. As part of an international MUZIC-2 programme (Mechanistic Understanding of Zr Corrosion and Hydrogen pickup), I have used a range of advanced microscopy techniques to study the microstructure, the nanoscale chemistry and the porosity in a series of zirconium alloys at different stages of corrosion and hydrogen pickup. Samples from both autoclave and in-reactor conditions were available to compare, I have focussed on RXA (recrystallised 580°C) Zr-1.0Nb and annealed (720°C) Zr-1.0Nb alloys. A set of samples from different exposures times were chosen to represent early, pre-transition and post-transition samples in order to compare the microstructure and microchemistry of the oxides, the metal-oxide interface and the metal. (Scanning) Transmission Electron Microscopy ((S)TEM), Transmission Kikuchi Diffraction (TKD) and automated crystal orientation mapping with TEM (ASTAR mapping) were used to study the grain structure and phase distribution. Significant differences in grain morphology were observed between samples oxidised in the autoclave with different corrosion rates, with more uneven metal-oxide interface, more parallel cracks and less organised oxide grains in the fast corroding samples. Comparing with autoclave samples, the in-reactor samples have shorter, less well-aligned monoclinic grains and more tetragonal grains. The rapidly oxidising annealed Zr-1.0Nb alloy also have much higher tetragonal grain fraction comparing with the slow corrosion rate RXA Zr-1.0Nb alloys. Porosity in the oxide is predicted to have a major influence on the overall rate of corrosion and hydrogen pickup, and there is much more porosity in the annealed Zr-1.0Nb alloy than found in either the RXA alloy or the similar alloy exposed to neutron irradiation. A combination of Energy Dispersion X-ray (EDX) mapping in STEM and Electron Energy Loss Spectroscopy (EELS) analysis of second phase particles can reveal the main and the minor element distributions respectively. The annealed Zr-1.0Nb alloys have Î2-Zr SPPs with nano crystalline structure and much larger size. Although they does not relate with the higher density of cracks in the oxide, the large SPP size can connect together all the small cracks that are generated by the huge amount of tetragonal to monoclinic phase transformation during corrosion and provides pathway for corrosion and hydrogen pickup. Two kinds of SPPs are found in the RXA Zr-1.0Nb alloys, one is Î2-Nb and another one is Zr-Nb-Fe Laves phase. Neutron irradiation seems to have little effect on promoting fast oxidation or dissolution of Î2-Nb precipitates, but encourages dissolution of Fe from Laves phase precipitates. Electron Energy Loss Spectroscopy (EELS) analysis of the oxidation state of Nb in Î2-Nb SPPs in the oxide revealed the fully oxidised Nb5+ state in the SPPs deep into the oxide, but Nb2+ in the crystalline SPPs near the metal-oxide interface. EELS, TKD and ASTAR mapping have also revealed the presence of suboxide layers with the hexagonal ZrO structure predicted by ab initio modelling. The combined thickness of the ZrO suboxide and oxygen-saturated layers at the metal-oxide interface correlates well to the estimated instantaneous oxidation rate, suggesting that the presence of this oxygen- rich zone combining with the part where porosity is not interconnected is the protective oxide that is rate limiting in the key in the transport processes involved in corrosion and hydrogen pickup.

Primary water stress corrosion cracking (SCC) of metals in pressurized water reactors (PWRs) is known to be one of the most challenging and cost intensive modes of failure in the nuclear industry. Even though it is known that cracking in Ni-base alloys proceeds mainly intergranular (IG), the initiation and propagation of cracks in ductile metals are not yet understood and a much-desired accurate prediction of SCC related failure seems unobtainable. In this thesis, a combination of microcantilever fracture experiments, scanning electron- (SEM) and transmission electron microscopy (TEM) techniques was employed to study and compare the failure of oxidized grain boundaries of Ni-base Alloy 600 with high and low intergranular carbide coverage and different sample history. A new technique for lifting-out whole cantilevers after testing and for performing 3D focussed ion beam sequencing (3D FIB-SEM) while preserving a thin central region of the cantilever for further TEM sample preparation was developed and is presented. In lieu with recent efforts of the main project sponsor Électricité de France (EDF) to build a predictive model for IGSCC based on localized/microscopic information, one of the main objectives was the extraction of the stress at failure of individual oxidized GBs. Supported by finite element simulations, microcantilever fracture tests revealed that surface oxides on top of individual GBs have the capability to alter the mechanical response by delaying/suppressing the onset of failure. An overestimation of the failure stress (&GT; 230 MPa) was observed, proving that the presence of the surface oxide on top of the test structures cannot be neglected. The failure stress on both samples, tested without influence of the surface oxide, was found to cover a range of 300 - 600 MPa, which agreed well with finite element simulations of the tests and further demonstrates the reliability of the obtained data. The second objective was to gain a better understanding of the observed fracture behaviour and the role of local microstructure. Using the gathered microscopy data, it was found that the crack clearly favours a progression along the IG oxide-metal interface in the presence of carbide precipitates. Electron energy loss spectroscopy (EELS) revealed that the observed crack path can be linked to compositional and density variations of the IG oxide. In the presence of carbides the oxide was layered. An oxide close to the stoichiometry of chromia was located at the original GB and next to the carbides. Next to this Cr-rich oxide, Fe-rich mixed spinel oxides of varying composition and density were found. An explanation for density variations based on the possible formation of defective spinel oxides of the type A2+B3+2O4, due to an unavailability of certain cation species is presented. No clear interface preference was observed in the absence of precipitation, where the IG oxide was found to be thin and often incomplete with Cr-richer oxides preferentially located at the original GB. While these observations were consistent on both samples (high and low carbide coverage), bigger void-like defects were located at the Fe-richer oxide-metal interface of the cold worked sample with high IG carbide precipitation only. These weak spots seemed to be the preferred path for crack propagation on this sample. The sample with low intergranular carbide coverage showed no obvious porosities at this interface but a Cr- depleted region was seen. Introducing a multi-faceted investigation strategy, supported by finite element simulations, the presented thesis provides the most accurate determination of the failure stress of oxidized GBs on Alloy 600 to date and and adds new valuable insights to our understanding of IGSCC and the future prediction of SCC related failures.

Advanced fission and fusion power offer long-term energy production from fundamentally carbon-free fuel sources, with high-energy densities and high reliability. However, for these power sources to be realised, many materials challenges need to be addressed. Oxide dispersion strengthened (ODS) ferritic alloys are recognized as potential fission fuel cladding and fusion structural materials because of their high-temperature stability and swelling-resistance imparted by the nano-scale oxide clusters they contain. This work presents the study of a set of model ODS alloys with a systematic variation in alloy chemistry in order to investigate the influence of chemical additions on the oxide cluster nature and the irradiation resistance. To investigate these properties, the high-resolution characterisation techniques of atom probe tomography and transmission electron microscopy were applied in combination with nanoindentation to assess mechanical properties and X-ray diffraction to study the crystallinity. The quantification of the character of the oxide particles in these ODS alloys by atom probe tomography presented a number of challenges and I extend the conventional analysis techniques in order to overcome these difficulties. By comparison of the addition of Ti, and then Cr, to a Fe-Y2O3 alloy, the influence of alloy chemistry on the oxide particles was evaluated. The main aims were to investigate the microstructure, the oxide cluster composition and to set a baseline to which the irradiated samples were compared. The irradiation damage of the oxide clusters was investigated using ion irradiation as an analogue for neutron damage. Irradiation experiments were performed at a range of temperatures and the stability of the clusters was assessed based on their size, number density and chemistry. Subtle changes in cluster chemistry at elevated temperatures and the dramatic changes which occurred after irradiation at low temperatures lead to conclusions about the competition of thermal diffusion and irradiation damage.