It is proposed to continue operating the graphite moderated Advanced Gas-cooled Reactor (AGR) fleet past its design life. Nuclear graphite's properties change in reactor and our limited mechanistic understanding of the relationship between graphite structure, across different lengthscales, and its properties limits our ability to predict its future behaviour. An improved understanding of the relationship between graphite's structural features, the relationship between features across different lengthscales and their effect on material properties would all contribute to a mechanistic understanding of graphite behaviour. Thermal expansion generates thermal strains and stresses in the graphite core during thermal transients, such as during reactor start-up and shut-down. Thermal expansion is a function of graphite crystal thermal expansion, crystallographic preferred orientation and microstructure, although the exact relationship between these is not understood. It is also altered by neutron irradiation. This thesis investigates graphite microstructure, virgin and irradiated, and its crystallographic preferred orientation, specifically as they pertain to thermal expansion. The microstructure of British nuclear graphites PGA and Gilsocarbon, used in the Magnox and AGR fleet respectively, have been investigated using scanning electron microscopy (SEM). Trepanned AGR graphite, that is, graphite drilled from the reactor brick during routine inspection is examined. These samples are from the 2012 Hinkley Point B inspection campaign and are taken from several points through the brick thickness. This provides a 'snap shot' of current AGR graphite condition. Deep trepan samples removed from further into the brick thickness are observed for the first time. Neutron damage was observed in Magnox graphite, irradiated in an inert environment in the material test reactor programme INEEL. The spatial variation in texture of PGA and Gilsocarbon, and the change in such texture after prestress was observed using synchrotron x-ray diffraction. Numerical models were used to identify the required texture change to produce changes in CTE, observed by other authors, during in-situ stress. PGA filler lamellae are arranged in parallel arrays and Gilsocarbon's smaller platelets are arranged in bunched clusters. Severe radiolytic oxidation is observed at all trepan locations, with oxidation decreasing away from the fuel. Radiolytic oxidation occurs at platelet edges. Texture measurements have indicated that PGA graphite exhibits significant spatial variation in texture. Gilsocarbon exhibits less variation but the variation observed is large enough to cause increased thermal stresses. Texture measurements of prestressed graphite have indicated that texture changes also vary spatially. Texture results and SEM observations indicate that spatial variation in texture is caused by spatial variation in microstructure. Changes to the filler particle during prestress may alter local texture. These results indicate there is a link between nuclear graphite's microstructure and its texture. The texture, a function of lamellae or platelet arrangement, determines its thermal expansion. Spatial variations in microstructure formed during manufacturing leads to spatial variations in CTE and possibly other texture sensitive properties, such as dimensional change. Deformation of the lamellae or platelets during stress; thermal creep or irradiation creep is expected to contribute to the observed change in properties associated with these stimuli.

Management of contamination from industrial activities and wastes from nuclear power generation and weapons development are arguably amongst the greatest challenges facing humanity currently and into the future. Understanding the mobility of toxic radioactive elements is essential for successful remediation strategies and safe management of our nuclear waste legacy (DEFRA, 2008). Interactions between minerals and radionuclides, such as sorption and precipitation, govern the mobility of the contaminants through the subsurface environment. Microbial metabolic processes (redox cycling or release of metabolites) have the potential to affect drastically these abiotic interactions. Microbially-driven mineralisation processes could provide long-term solid-phase-capture solutions to radionuclide contamination problems and support safety cases for geological disposal of radioactive waste. The recent advancements at the intersection between mineralogy, microbiology and radiochemistry were reviewed with the aid of a cluster analysis (Self-Organising Map). This is a relatively novel method of creating a map of the ‘research landscape’ which provides a visual summary of the reviewed literature and can help to identify areas of promising and active research as well as less researched interdisciplinary areas. It is the first time this tool has been applied to research literature on this interdisciplinary topic, and it highlighted the need to gain further understanding of ternary systems including bacteria, minerals and radionuclides. The analysis showed that phyllosilicates are of interest, but few studies have explored the properties of the Fe(II)/Fe(III)-containing micas biotite and chlorite. The ability of model Fe(III)-reducing microorganisms to reduce Fe(III) in biotite and chlorite was demonstrated in batch model systems. In chlorite, approximately 20% and in biotite ~40% of the bulk Fe(III) was transformed to Fe(II) by this reduction. To our knowledge, this is the first study to show the availability of Fe(III) in biotite for such reduction and the ability of the model organism Shewanella oneidensis MR-1 to conserve energy for growth using Fe(III) in biotite as the sole electron acceptor. The microbial Fe(III) reduction led to a decrease in the sorption of Cs and Sr by chlorite, but had very little effect on sorption to biotite. The data indicate that remediation strategies based on microbial Fe(III) reduction may exacerbate the movement of Cs and Sr through strata where sorption is dominated by phyllosilicates, particularly chlorite. While microbial Fe(III) reduction had only a slight effect on the sorption properties of biotite and chlorite, it drastically altered their redox properties. Previously bioreduced biotite and chlorite readily removed Cr(VI), Tc(VII) and Np(V) by surface-mediated reduction. The minerals were also able to reduce U(VI), but solution chemistry affected this reaction, reflecting the complexity of the biogeochemistry of this actinide. Overall, this work highlights the importance of decoupling microbial and geochemical processes in developing a holistic understanding of radionuclide behaviour in the environment. This body of work forms the thesis is entitled ‘Mineralisation and Biomineralisation of radionuclides’, and was prepared by Diana Roumenova Brookshaw for submission in August 2013 for the degree of Doctor of Philosophy to the University of Manchester.

If new build nuclear reactors are built in the UK they will provide a large low carbon thermal resource that can be recovered for heating services through heat networks (district heating). There are however questions about the geographic location of nuclear sites relative to heating demand and public/user interpretations of a potentially controversial technology to consider. This thesis includes three research themes that explore these issues. The first is an assessment of potential non-technical barriers to nuclear heat network development. The second is a focus group approach to studying local resident responses to nuclear heat network technology both as potential users, and as public groups. The third theme considers the technical potential for a heat network connecting the Hartlepool nuclear site to local heating demand centres. The research finds that there is potential for nuclear heat networks to take 70,000 existing users off the natural gas in the Hartlepool area. Following series of expert interviews it finds no non-technical barriers that would be unique to nuclear heat networks as opposed to other heat network types. It also suggests that the technology could be acceptable to local residents if it is framed as a local resource that benefits the local area. These findings indicate that there could be similar potential at Heysham and Oldbury nuclear sites.

The treatment of advanced nuclear fuels is relevant to the stabilisation of legacy spent fuels or nuclear materials and fuels from future nuclear reactors. Historically, spent fuel reprocessing has been driven to recover uranium and plutonium for reuse. Future fuel cycles may also recover the minor actinides neptunium, americium and perhaps curium. These actinides would be fabricated into new reactor fuel to produce energy and for transmutation of the minor actinides. This has the potential to reduce the long lived radioactivity of the spent fuel and reprocessing high level waste, whilst also maximising energy production. To achieve these aims there are a range of materials that could be used as advanced nuclear fuels, these include metals, oxides, carbides, nitrides and composite materials, and these fuels may also be alloyed. These advanced fuels may need to be reprocessed, and as head end is the first chemical treatment step in a reprocessing plant, the issues caused by treating these advanced fuels are faced primarily by head end. Changes to the overall reprocessing specification, such as reduction in discharge authorisations for volatile radionuclides, will have the greatest impact upon head end. All these factors may lead to the introduction of pre-treatment technologies (e.g. Voloxidation) or enhanced dissolution technologies, e.g. mediated dissolution using silver(II).Literature and experimental studies show that uranium dioxide and low plutonium content MOx dissolves in nitric acid via direct and indirect nitrate reduction. The indirect nitrous acid catalysed route is kinetically most significant. The kinetics for the dissolution of uranium dioxide and 5 % plutonium MOx have been derived experimentally. Studies of the dissolution of MOx pellets in concentrated nitric acid and near boiling conditions indicate that dissolution shows a degree of mass transfer limitation. Thermodynamic studies show that the pronounced reduction in the MOx dissolution extent at 30-40% plutonium is due to the thermodynamics of the key dissolution reactions. One technology that could be used to dissolve plutonium-rich residues that are generated from the reprocessing of MOx fuels is mediated dissolution. Inactive studies using linear staircase voltammetry (LSCV) and constant current bulk electrolysis (BE) have been used to optimise a 100 ml dissolution cell. The generation of silver(II) is dependent upon silver concentration, agitation and the size of the separator membrane. Whilst the stability of silver(II) is defined by the kinetics of water oxidation, this is dependent upon a number of factors including nitric acid concentration, silver(I):(II) ratio, temperature and the rate of migration from the catholyte into the anolyte. LSCV experiments have shown that Tafel analysis confirms there is a good relationship between potential and anode current density assuming oxygen evolution and silver(I) oxidation. Kinetic modelling of the BE experiments can be used to model the silver(II) generation, steady state and decomposition due to reaction with water. The dissolution cell has been demonstrated to be capable of dissolving plutonium dioxide to 200 g.l-1 in less than 2 hours with good faradaic efficiency.

The University of Manchester Greg BlackThesis submitted for the degree of Doctor of EngineeringIrradiated Graphite Waste: Analysis and Modelling of Radionuclide Production with a View to Long Term Disposal23rd June 2014The UK has predominantly used graphite moderator reactor designs in both its research and civil nuclear programmes. This material will become activated during operation and, once all reactors are shutdown, will represent a waste legacy of 96,000 tonnes [1]. The safe and effective management of this material will require a full understanding of the final radiological inventory. The activity is known to arise from impurities present in the graphite at start of life as well as from contamination products transported from other components in the reactor circuit. The process is further complicated by radiolytic oxidation which leads to considerable weightloss of the graphite components. A comprehensive modelling methodology has been developed and validated to estimate the activity of the principle radionuclides of concern, 3H, 14C, 36Cl and 60Co. This methodology involves the simulation of neutron flux using the reactor physics code WIMS, and radiation transport code MCBEND. Activation calculations have been performed using the neutron activation software FISPACT. The final methodology developed allows full consideration of all processes which may contribute to the final radiological inventory of the material. The final activity and production pathway of each radionuclide has been researched in depth, as well as operational parameters such as the effect of changes in flux, fuel burnup, graphite weightloss and irradiation time. Methods to experimentally determine the activity, and distribution of key radionuclides within irradiated graphite samples have been developed in this research using a combination of both gamma spectroscopy and autoradiography. This work has been externally validated and provides confidence in the accuracy of the final modelling predictions. This work has been undertaken as part of the EU FP7 EURATOM Project: CARBOWASTE, and was funded by the Office for Nuclear Regulation.

Les interactions plasma-paroi représentent l'un des principaux problèmes à résoudre pour avancer dans la recherche sur la fusion contrôlée. Ce travail de thèse a pour objectif de développer une source d'hydrogène atomique à basse pression (< Torr) dans un réacteur à configuration hélicon en mélange H2/Ar pour étudier la gravure chimique du graphite et de composites à fibres de carbone utilisés dans le tokamak Tore Supra. Selon les conditions expérimentales, le réacteur peut générer les modes de couplages capacitif, inductif, Trivelpiece-Gould et hélicon à bas champ. Leur caractérisation a montré que le mode inductif est, dans ce cas, celui présentant le plus grand intérêt pour la source d'hydrogène atomique. Les études en mode inductif ont révélé un phénomène de décroissance de la densité relative de deux niveaux métastables de l'ion Ar+ et d'un niveau métastable de l'argon neutre lors de l'augmentation du champ magnétique de confinement. Un modèle simple a confirmé que ces niveaux métastables sont détruits par collisions électroniques vers des niveaux de plus grande énergie. La gravure du graphite par la source d'hydrogène atomique est relativement efficace (jusqu'à 3 µm/h) à 10 mTorr et diminue avec la pression. Une analyse qualitative de la cinétique de l'hydrogène atomique a permis de conclure que cette baisse de la vitesse de gravure est due au flux d'hydrogène atomique arrivant sur l'échantillon qui décroît lorsque la pression augmente. Les premières observations de la surface après gravure ont mis en évidence la présence de structures carbonées (agglomérats de nanoparticules et dépôts). Ces structures ressemblent à celles observées dans différents tokamaks / Plasma-wall interactions are one of the main issues in fusion research and must be thoroughly studied to progress in this topic. The objective of this work is to develop an atomic hydrogen source at low pressure (< Torr) in a helicon configuration reactor working in H2/Ar gas mixture. This source is then used to study the chemical etching of graphite and carbon fiber composites composing the limiter of the Tore Supra tokamak. Depending on the experimental conditions, the RF power coupling of the reactor can be in capacitive, inductive, Trivelpiece-Gould or low field helicon mode. The characterization of these modes determined that in this case the inductive one presents the greatest interest for the atomic hydrogen source. Further studies in inductive mode showed that increasing the confinement magnetic field leads to a decrease of measured relative densities of two metastable levels of argon ion and one metastable state of neutral argon. A simple model simulating neutral metastable state behavior confirmed that these levels are destroyed by electronic collisions towards upper levels. The chemical etching of graphite exposed to the atomic hydrogen source is relatively efficient (up to 3 µm/h) at 10 mTorr and drops with the pressure. A qualitative analysis of atomic hydrogen kinetics concluded that this behavior is due to the decrease of atomic hydrogen flux on the sample with increasing pressure. Finally, first observations of the etched surface underlined different structures (nanoparticles clusters and deposits). These can be compared to the ones observed in different tokamaks

La nécessité de trouver une nouvelle source d'énergie a mené les scientifiques à explorer la voie de la fusion thermonucléaire par confinement magnétique. Cependant la réalisation de tels plasmas de fusion dans les tokamaks actuels pose de nombreux défis tels que les interactions entre le plasma et les parois à l'origine de la création de poussières pouvant être néfastes au bon fonctionnement des futurs réacteurs à fusion. Une bonne connaissance de la quantité de poussières produites, de leur localisation et de leur transport durant la phase plasma est donc d'une importance fondamentale pour l'exploitation d'ITER. Un algorithme, développé et validé par l'expérience, est utilisé pour détecter et suivre les poussières dans ASDEX Upgrade (AUG) durant la phase plasma. Il permet d'analyser automatiquement des vidéos enregistrées par caméras rapides. Une large statistique sur la quantité de poussières micrométriques détectées en fonction du temps cumulé de décharge plasma est réalisée. Les premières analyses effectuées sur les cinq dernières campagnes montrent que la quantité de poussières est significativement faible voire nulle dans la plupart des décharges effectuées dans AUG, excepté pour des conditions spécifiques de décharges correspondant à des phases anormales de fonctionnement (disruptions, ELMs, déplacements du plasma vers les CFPs et absorption inefficace de la puissance de chauffage). Ces observations par caméra rapide et l'analyse via l'algorithme peuvent ainsi permettre, avec l'utilisation d'autres diagnostics plasmas, d'identifier les décharges plasmas à risque, pouvant aider à sélectionner les scénarios de fonctionnement les plus efficaces pour ITER / The necessity to find a new energy source has lead scientists to explore the way of thermonuclear fusion by magnetic confinement considered as one of the most promising possibility. However the production of such plasmas in the current tokamaks lies to several challenges like the interactions between the plasma and the first wall which spark off the creation of a lot of dust in the plasma which could be problematic for the operation of the next fusion reactors. The knowledge of dust production rates, localisation and transport through the vacuum vessel during plasma phases is of primary importance and must be investigated in preparation of ITER. A time and resource efficient algorithm named TRACE, validated thanks to a dedicated laboratory experiment, is used to detect and track dust particles in ASDEX Upgrade during plasma phase. It allows for automatically analyzing videos originating from fast framing cameras. A statistic about micron sized dust detection rate as a function of cumulated discharge duration is made on a large number of discharges (1470). First analyses covering five last campaigns clearly confirm that the amount of dust is significantly low in most of discharges realized in ASDEX Upgrade, excepted for specific conditions corresponding to off-normal plasma phases (disruptions, strong plasma fluctuations including ELMs, plasma displacement toward PFCs and inefficient absorption of heating power). These observations allow to identify the risky plasma discharges and choose the most efficient plasmas scenarios for ITER. It seems to also confirm the applicability of an all tungsten first wall for future fusion reactors as ITER

Cette étude s'inscrit dans la problématique de l'extraction minière de l'uranium. L'étape plus spécifiquement visée est celle de l'extraction liquide-liquide au cours de laquelle l'uranium en solution aqueuse de lixiviation est transféré dans une phase organique par une complexation avec une amine tertiaire (Alamine 336®). L'étude porte sur cette étape de séparation par le biais d'un contact liquide-liquide sans dispersion assuré au travers d'une membrane macroporeuse organique (pertraction). Cette technologie permet l'utilisation de phases de densités proches. Ainsi, l'extraction de l'uranium et du molybdène, co-extrait industriellement par l'Alamine 336, ont été étudiées pour différentes concentrations élevées en molécules extractantes. Pour cela, l'ensemble des paramètres physico-chimiques influents du système chimique (courbe de distribution, viscosité, densité, tension de surface, coefficient de diffusion à l'infini, etc.) et de la membrane (porosité, dimensions, tortuosité, etc.) sont caractérisés. Des essais de pertraction sur un dispositif unitaire de fibre creuse, développé dans le cadre de cette étude, sont réalisés et les résultats sont modélisés par une approche de résistances du transfert de matière en séries. Un seul paramètre ajustable est retenu : le coefficient de diffusion en phase organique. Cette modélisation permet de mettre en évidence les limitations au transfert de l'uranium de la phase aqueuse vers la phase organique lors du processus d'extraction à travers la membrane liées au système chimique, aux débits et à la membrane. / This study is falls within the framework of uranium mining. The step more specifically aimed is the solvent extraction during which the uranium is transferred from a lixiviation aqueous solution to an organic phase by a complexation with a tertiary amine (Alamine 336®). The study focuses on this step of separation using a liquid-liquid contact without dispersal guaranteed by an organic macroporous membrane (pertraction). This technology allows the use of phases with close densities. So, uranium and molybdenum extraction, co-extracted industrially by Alamine 336, were studied for various high extractantes molecules concentrations. For that purpose, all the influential physical chemical parameters of the chemical system (distribution curve, viscosity, density, surface tension, infinite diffusion coefficient, etc.) and of the membrane (porosity, size, tortuosity, etc.) are characterized. Pertraction essays on a single hollow fibre, developed within the framework of this study, are performed and the results are modelized by an approach of mass transfer resistances in series. A single adjustable parameter is retained: the diffusion coefficient in organic phase. This modelling allows highlighting the limitations of uranium transfer from the aqueous phase towards the organic phase during the extraction process through the membrane function of chemical system, the flows and the membrane.

The work reported in this thesis explores innovative control structures and controller design for a heavy duty Caterpillar C6.6 diesel engine. The aim of the work is not only to demonstrate the optimisation of engine performance in terms of fuel consumption, NOx and soot emissions, but also to explore ways to reduce lengthy calibration time and its associated high costs. The test engine is equipped with high pressure exhaust gas recirculation (EGR) and a variable geometry turbocharger (VGT). Consequently, there are two principal inputs in the air-path: EGR valve position and VGT vane position. The fuel injection system is common rail, with injectors electrically actuated and includes a multi-pulse injection mode. With two-pulse injection mode, there are as many as five control variables in the fuel-path needing to be adjusted for different engine operating conditions.

Graphite is used as a moderating material and as a structural component in a number of current generation nuclear reactors. During reactor operation stresses develop in the graphite components, causing them to deform. If significant numbers of graphite components were to fail in this manner, the material’s effectiveness as a neutron moderator will be reduced, and the reactor’s safe operation may be compromised. It is therefore important to understand how the microstructure of graphite affects the material’s response to these stresses. Despite much research into the effects of stress on nuclear grade graphite, there remain gaps in our understanding of this process, and there are a number of frequently observed limitations in the current research. Many existing studies either focus on the bulk material, ignoring the important changes at the microlevel; or focus on residual stresses due to the lack of available in-situ data. An experimental programme was designed to study stress-induced changes to the microstructures of Gilsocarbon and Pile Grade A graphite used in UK nuclear reactors. Particular focus was paid to the deformation of the pore structure, since graphite is highly porous and the porosity has a significant effect on the strength and structural integrity of the graphite components. A compression rig was used to simulate the build-up of operational stresses, during which confocal laser microscopy and X-ray tomography were performed to quantify changes to the pore structure at the microlevel; while X-ray diffraction was performed to study deformation of the crystal lattice and quantify the build-up of lattice strains. Pore properties of interest included pore area, surface area, volume, eccentricity, orientation, angularity and separation. Crystal lattice properties of interest included layer spacing, unit cell and crystallite size parameters, lattice strains and Bacon Anisotropy Factor. The experimental and analytical techniques were designed to significantly enhance our current understanding of how graphite responds to stress, with each observation made using a novel technique or improving the effectiveness of existing techniques. These studies have enabled significant novel observations and discussions of the stress-induced deformation behaviour of nuclear grade graphite to be made.