The aim of this thesis is to provide insight into the structures and properties of the uranium oxides. As UO2 is easily oxidised during the nuclear fuel cycle it is important to have a detailed understanding of the structures and properties of the oxidation products. Experimental work over the years has revealed many stable oxides including UO2, U4O9, U3O7, U2O5, U3O8 and UO3, all with a number of different polymorphs. The oxides are broadly split into two categories, fluorite-based structures with stoichiometries in the range of UO2 to U2O5 and less dense layered-type structures with stoichiometries in the range of U2O5 to UO3. While UO2 is well characterised, both experimentally and computationally, there is a paucity of data concerning higher stoichiometry oxides in the literature. Experiments and simulations are emerging that deal with individual phases, however a comprehensive study that assesses the properties of all polymorphs and provides comparison over the full range of stoichiometries has been lacking from the literature First the nuclear fuel cycle is introduced, as well as UO2 as a nuclear fuel (Chapter 1), before the quantum mechanical methodology used throughout is explained (Chapter 2). Applying a number of different density functionals (including GGAs, meta-GGAs and hybrids) to UO2 in Chapter 3 it emerges that the PBE + U formalism reproduces the experimentally observed properties to a good degree of accuracy, and so is selected for the rest of the simulations. Following this Chapter 4 examines defect clusters in UO2, finding split interstitials to dominate at low stoichiometry (UO2 – UO2.0625), chains of 2:2:2 Willis clusters at higher stoichiometry (UO2.125 – UO2.25 (U4O9)) and split quad interstitials at higher stoichiometry (UO2.33 (U3O7)). Chapter 5 is an investigation of layered U2O5, where it emerges that the Np2O5 structure is more stable than δ-U2O5 and all uranium ions are in the U5+ oxidation state. Next Chapter 6 considers layered U3O8, which is structurally oxygen rich U2O5, where it is found that U5+ and U6+ ions exist in pentagonal bipyramidal and octahedral coordination respectively. The final set of results in Chapter 7 concern the polymorphs of UO3, where it is found that U6+ adopts a range of coordination environments and the predicted relative stability of each modification matches well with experiment. Finally the conclusions are presented in Chapter 8 along with plans for future work.

abstract: ABSTRACT The behavior of the fission products, as they are released from fission events during nuclear reaction, plays an important role in nuclear fuel performance. Fission product release can occur through grain boundary (GB) at low burnups; therefore, this study simulates the mass transport of fission gases in a 2-D GB network to look into the effects of GB characteristics on this phenomenon, with emphasis on conditions that can lead to percolation. A finite element model was created based on the microstructure of a depleted UO2 sample characterized by Electron Backscattering Diffraction (EBSD). The GBs were categorized into high (D2), low (D1) and bulk diffusivity (Dbulk) based on their misorientation angles and Coincident Site Lattice (CSL) types. The simulation was run using different diffusivity ratios (D2/Dbulk) ranging from 1 to 10^8. The model was set up in three ways: constant temperature case, temperature gradient effects and window methods that mimic the environments in a Light Water Reactor (LWR). In general, the formation of percolation paths was observed at a ratio higher than 10^4 in the measured GB network, which had a 68% fraction of high diffusivity GBs. The presence of temperature gradient created an uneven concentration distribution and decreased the overall mass flux. Finally, radial temperature and fission gas concentration profiles were obtained for a fuel pellet in operation using an approximate 1-D model. The 100 µm long microstructurally explicit model was used to simulate, to the scale of a real UO2 pellet, the mass transport at different radial positions, with boundary conditions obtained from the profiles. Stronger percolation effects were observed at the intermediate and periphery position of the pellet. The results also showed that highest mass flux happens at the edge of a pellet at steady state to accommodate for the sharp concentration drop. / Dissertation/Thesis / M.S. Materials Science and Engineering 2011

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.

No description available.

Nuclear energy has gained crucial importance since it has a minor impact on climate change and greenhouse gas releases; additionally, the other energy sources are insufficient to reach the world's energy needs without nuclear energy. Another sign that the Generation IV International Forum (Kelly, Gen IV International Forum: A decade of progress through international cooperation, 2014) has pointed out is to utilize uranium resources to the maximum and recycle spent nuclear fuel through burn-up in the Generation IV reactor designs, one of which is the molten salt reactor (MSR). Therefore, the MSR can use the spent nuclear fuel as a fresh fuel when the actinides recycle. That reprocessing of spent fuel could be one of the opportunities to contribute to future nuclear energy goals. This study aims to develop a modified pyroprocessing method to prepare molten salt fuels for MSR from spent oxide nuclear fuel that was burned in light water reactors (LWRs). The process diagram illustrated as (1) spent fuel treatment, (2) chopping and voloxidation of spent oxide fuel, (3) oxide reduction of spent fuel, and then depending on the fuel structure and composition for the MSR, it continues by one or two of the following; – electrorefining, – chlorination, and – fluorination. The subject of this study focused on oxide reduction in two categories: chemical reduction and electrochemical reduction. The system designs have been optimized in calcium salts since they have high calcium metal and calcium oxide solubility. The significant results indicated that both methods would substantially reduce the solid uranium dioxide pellet. The chemical reduction will reduce the total solid pellet at 850oC in the composition of 55.73mol%CaCl2-12.37mol%CaF2-26.58mol%Ca-5.32mol%UO2 over 12 hours. The total reduction in the electrochemical test is seen at 850oC during 12 hours with a salt composition of 79mol%CaCl2-17mol%CaF2-4mol%CaO. These oxide reduction mechanisms are convenient ways to reprocess spent oxide fuel from LWRs to utilize in the MSR. Additionally, the reduced fuel is also applicable to using other next-generation reactors. The prospect of this research is the explicit comparison between chemical and electrochemical methods in calcium salts. / M.S. / Nuclear energy is a crucial energy production to meet the world’s future energy needs. The 6 (six) next-generation reactor design has been determined based on their sustainability, economic, and peaceful application for the world. One of those designs is molten salt reactors (MSRs) which have more attention due to their fuel choice. Most MSRs use the reprocessed fuel from current reactors or the fuel with the breeder blanket that creates more fuel while the reactor operates. This study aims to provide a diagram showing the various steps involved in the preparation of molten salt fuel from spent oxide fuel, which is a mainly utilized form of fuel in current and previous operations. The flowsheet’s first step is the treatment of spent fuel that releases most of the decay heat. The second step is that spent fuel chopping and voloxidation, which meets the requirements of removing gas products and cladding material from used fuel. Afterward, the spent oxide fuel reduces into its metal form chemically or electrochemically in oxide reduction. Then, the molten salt fuel could be fabricated in n one or two more steps from reduced metals: electrorefining, chlorination, or fluorination. Chlorination and fluorination pass through the specific gas components to convert the metal forms into salt. Electrorefining could be applied to arrange the composition of the reduced metal, and this stage is strongly dependent on the MSR designs; it may get eliminated due to its unnecessity. The oxide mechanisms mentioned above were examined under different design conditions to acquire a total reduction of the fuel pellet in calcium salts. The chemical reduction and electroreduction experiments have shown the reduced whole pellet at 850oC with two different salt mixtures. The design impacts of the reduction mechanism were discussed extensively between chemical and electrochemical reductions to identify the benefits and limitations.

Made available in DSpace on 2014-10-09T12:35:27Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:05:40Z (GMT). No. of bitstreams: 0 / Dissertação (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Made available in DSpace on 2014-10-09T12:35:27Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:05:40Z (GMT). No. of bitstreams: 0 / No Brasil e no mundo a tecnologia nuclear vem ocupando posição de destaque com diversas aplicações na indústria, geração de energia, meio ambiente e na medicina, melhorando a qualidade de exames e tratamentos, consequentemente, a vida das pessoas. O urânio é o principal elemento utilizado em instalações nucleares, servindo como material base desde a geração de eletricidade à fabricação de radiofármacos. Nos anos 50, em meio à guerra fria, a então recém-criada Agência Internacional de Energia Atômica se propôs a supervisionar instalações nucleares e incentivar a fabricação de combustíveis nucleares com baixo teor de urânio, conhecidos como combustíveis do tipo Material Test Reactor (MTR), fabricados inicialmente na forma de U3O8 e mais tarde o U3Si2, ambos dispersos em alumínio. A utilização desta tecnologia requer uma constante melhoria de todos os processos que envolvem a fabricação do MTR sujeita a diversos protocolos internacionais, os quais procuram garantir a confiabilidade desse combustível do ponto de vista prático e ambiental. Dentro desse contexto, o controle de impurezas, do ponto de vista da economia de nêutrons, afeta diretamente a qualidade de qualquer combustível nuclear, fazendo-se necessário um controle rigoroso. A literatura reporta procedimentos que, além de gerar resíduos, são demorados e dispendiosos, pois necessitam de curva de calibração univariada e materiais de referência. Assim, o objetivo deste trabalho é estabelecer e validar uma metodologia de análise química quantitativa não destrutiva, de baixo custo e tempo de análise, tal como, minimizar a geração de resíduo para a determinação multielementar dos maiores constituintes (Utotal e Si) e as impurezas (B, Mg, Al, Cr, Mn, Fe, Co, Ni, Cu, Zn, Mo, Cd e outros) presentes em U3O8 e U3Si2, atendendo as necessidades de reatores nucleares na qualificação de combustíveis nucleares do tipo MTR. Para tanto, foi aplicada a técnica de fluorescência de raios X que permite análises químicas rápidas e não destrutivas, além de não necessitar de tratamentos químicos prévios (dissolução, digestão e outros) na fase de preparação de amostras. Para as correções de efeitos espectrais e de matriz foram aplicados e avaliados os métodos de parâmetros fundamentais, de curva de calibração univariada e de calibração multivariada. Os resultados foram comparados por meios de testes estatísticos em conformidade com a norma ISO 17025 com os MRCs (123(1-7) e 124(1-7)) de U3O8 da New Brunswick Laboratory (NBL) e 16 amostras de U3Si2 cedidas pelo CCN do IPEN-CNEN-SP. A quimiometria demonstrou-se um método promissor para a determinação de maiores e menores constituintes em combustíveis nuclear a base de U3O8 e U3Si2, uma vez que a precisão e exatidão são estatisticamente iguais aos métodos de análises volumétrica, gravimétrica e ICP-OES. / Dissertação (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

L’objectif de cette thèse est d’étudier et de reproduire les caractéristiques spécifiques de la microstructure du combustible nucléaire irradié à fort taux de combustion, appelée structure HBS (High Burnup Structure). Il s’agit d’étudier les différents paramètres pertinents impliqués dans la formation d’une telle structure, en évaluant leur importance, et en clarifiant leurs éventuelles synergies. Cet objectif a été réalisé en utilisant un système de modèle ultra simplifié, à savoir des monocristaux de dioxyde d’uranium (UO₂) irradiés par des ions de basse énergie (quelques centaines de keV) de Lanthane (La) ou de xénon (Xe) à une température de 773 K, correspondant à celle de la périphérie des véritables pastilles de combustible en réacteur. Les énergies et les masses des ions ont été choisies pour étudier la déstabilisation du solide en fonction de deux paramètres-clefs: (i) les collisions nucléaires élastiques et (ii) la contribution chimique de l'incorporation d'impuretés à forte concentration. Les deux espèces ont été choisies délibérément pour leurs solubilités très différentes dans le dioxyde d’uranium: les ions La sont solubles dans l'UO₂ jusqu’à de très fortes concentrations, tandis que les ions Xe sont insolubles. Les techniques de la Microscopie Électronique en Transmission (TEM) et de Spectrométrie de Rétrodiffusion Rutherford en canalisation RBS/C ont été conduites in situ couplée avec l’irradiation. Ces deux techniques utilisées pour visualiser, quantifier et fournir des informations concernant la fraction des défauts induits par l’irradiation et la formation des bulles, de cavités ou de précipités dans le solide. Les données de canalisation ont été analysées par simulation Monte-Carlo en supposant l’existence de deux catégories de défauts : (i) des atomes aléatoirement déplacés (RDA) et (ii) des distorsions des rangés atomiques (BC). L’évolution de la fraction de défaut de type RDA montre une forte augmentation entre 0.4 à 4.0 dpa (correspondant à une très faible concentration des ions implantés), indépendamment de la nature des ions. Elle est suivie par une saturation de la fraction de RDA pour les deux ions sur une large gamme d’irradiation quoi s’étend jusque 100 dpa. Une forte élévation de la fraction de RDA est observée en particulier pour les cristaux implantés avec des ions Xe pour une concentration élevée dépassant les 4%. En ce qui concerne l’évolution de BC, elle augmente fortement jusqu’à 4 dpa et sature ensuite deux ions La et Xe. Les résultats de microscopie électronique in situ montrent que des défauts identiques pour les deux ions induits par l’irradiation apparaissent, et présentent la même évolution en fonction de la fluence. Les différents défauts évoluent en fonction de la fluence : la première étape correspond à la formation de ‘black dots’ ; la deuxième étape est caractérisée par la formation de boucles puis de lignes de dislocations, qui évoluent finalement jusqu’à commencer à devenir moins différenciables; le processus de restructuration se poursuit et forme un réseau de dislocations enchevêtrées. Une forte densité de bulles de gaz, de taille nanométrique et avec un diamètre moyen de 2 nm est observée pour le cristal Xe implanté à une dose seuil de 4 dpa. Le couplage des deux techniques conduites in situ montre que la différence entre les valeurs à saturation des fractions RDA des deux ions, d’une part, et l'augmentation drastique de RDA à très forte concentration d'ions Xe implantés d’autre part peuvent être attribuées à : (i) la solubilité des ions La vis-à-vis des ions Xe, conduisant à la formation des bulles de gaz de taille nanométrique et (ii) la taille des espèces implantées dans la matrice UO₂, pour laquelle les atomes Xe insolubles ont un rayon atomique beaucoup plus grand que le rayon cationique des atomes U⁴⁺(les atomes La³⁺ ont un rayon atomique similaire à celui des atomes U⁴⁺), responsable de plus de contraintes supplémentaires dans la structure cristalline. / The aim of this thesis is to investigate and reproduce the specific features of the microstructure of the high burnup structure of the irradiated nuclear fuel and to explore the various relevant parameters involved in the formation of such a structure, in evaluating their importance, and in clarifying the synergies between them. This have been performed by using a very simplified model system – namely uranium dioxide single crystals- irradiated with low energy La and Xe ions at 773 K, corresponding to the temperature at the periphery of the genuine fuel. The energies and masses of bombarding ions were chosen to investigate the destabilization of the solid due to: (i) the elastic nuclear collisions and by (ii) the chemical contribution of implanting impurities at high concentrations by implanting different ions in UO₂, namely Xe and La, having very different solubility: La species are soluble in UO₂ while Xe ions are insoluble. In situ Transmission electron Microscopy (TEM) and in situ Rutherford Backscattering Spectrometry in the channeling mode (RBS/C), both techniques coupled to ion irradiation, were performed to visualize, quantify and provide information with respect to the fraction of radiation-induced defects and the formation of bubbles, cavities, or precipitates. The channeling data were analyzed afterwards by Monte Carlo simulations assuming two class of defects comprising (i) randomly displaced atoms (RDA) and (ii) bent channels (BC) defects. Regarding the RDA evolution, a sharp increase step appears from 0.4 to 4.0 dpa (corresponding to a low concentration of implanted ions) regardless of nature of ions followed by a saturation of the fraction of RDA for both ions over a wide range of irradiation. A sharp increase of RDA fraction is observed specifically for crystals implanted with Xe ions at a high concentration exceeding 1.5 % (corresponding to the dose of more than 125 dpa). Regarding the BC evolution, for both ions, the evolution shows an increase in the fraction of BC up to 4.0 dpa then the fraction of BC almost saturates for Xe and La ions. In situ TEM results show that a similar radiation-induced defects appear for both ions and the same evolution of defects as a function of the fluence is observed. The various defects evolved as a function of the fluence: starting from the black dot defects formation that were observed as a first type of defects created, then dislocation lines and loops appeared and evolved until they started to be become less distinguishable, the restructuring process continued by forming a tangled dislocation network. A high density of nanometer-sized gas bubbles with a mean diameter 2 nm were observed at room temperature for the Xe-implanted crystal at a threshold dose of 4 dpa. The coupling between both techniques (in situ RBS/C and TEM) demonstrates that the difference between the two plateaus of saturation between the two ions and the dramatic increase of RDA at high concentration of implanted Xe ions can be ascribed to: (i) the solubility of La compared to Xe ions leading to the formation of nanometer-sized gas bubbles and (ii) the size of implanted species in UO₂ matrix where insoluble Xe atoms have an atomic radius much larger than the cationic radius of U⁴⁺ atoms, (La³⁺ atoms have a similar atomic radius as U⁴⁺ atoms) responsible for more stress in UO₂ crystal.