Le procédé actuel de fabrication des combustibles nucléaires MOX (UO2-PuO2) est réalisé exclusivement par voie sèche (broyage-tamisage des poudres, pressage et frittage). Afin d’améliorer ce procédé, des recherches sont menées sur le développement d’un procédé de fabrication du MOX par voie liquide. Ce procédé vise à réduire l’empoussièrement des boîtes à gants, améliorer l’homogénéité U/Pu et diminuer la quantité de défauts (fissures, éclats) des pastilles frittées, afin de minimiser le taux de pastilles rebutées. Dans cette optique, le proceed voie liquide étudié consiste à préparer une suspension aqueuse à partir des poudres brutes d’oxydes (mélange d’eau, d’additifs organiques et de poudres), puis à lui faire subir une granulation. Les granules obtenus sont ensuite pressés et frittés. Le procédé de granulation étudié est un procédé innovant de granulation cryogénique, consistant à atomiser la suspension dans de l’azote liquide puis à lyophiliser les granules gelés ainsi formés. L’étape clé du procédé est la préparation de suspension aqueuse de poudres UO2-PuO2, qui doit être dispersée, homogène, stable et suffisamment fluide pour l’étape de granulation. Une étude préliminaire a été réalisée avec des poudres simulantes, choisies pour leurs propriétés en suspension : TiO2 et Y2O3 pour simuler respectivement UO2 et PuO2. Une fois maîtrisé avec les poudres simulantes, ce procédé a été étudié avec UO2 et PuO2 pour déterminer ses conditions optimales de mise en oeuvre. La dispersion de suspensions d’UO2 et/ou de PuO2 a été étudiée par acoustophorométrie et rhéologie afin d’optimiser leur formulation, ainsi que l’étape de granulation cryogénique. Enfin, des pastilles d’UO2 et d’UO2-PuO2 ont été formées à partir des granules, puis leur frittage et leur microstructure ont été étudiés. / The current manufacturing process for MOX nuclear fuels (UO2-PuO2) is carried out by dry route exclusively (grinding, sieving, pressing and sintering). In order to improve this process, research is conducted to develop a liquid route MOX manufacturing process. It aims to reduce glove boxes dusting, increase U/Pu homogeneity and reduce the amount of defects (cracks, voids) in sintered pellets, in order to minimize scraps. In this scope, the liquid process studied consists in the preparation of an aqueous suspension from the raw oxide powders (mix of water, organic additives and powders), which is then granulated. The resulting granules are pressed into pellets and sintered. The granulation process studied is an innovative freeze granulation process that implies to spray the suspension in liquid nitrogen and then freeze-dry the frozen granules that are formed.The key step of the process is the preparation of aqueous suspension of UO2-PuO2 powders, which must be dispersed, homogeneous, stable and fluid enough for the granulation step. A preliminary study was carried out using surrogate powders, chosen for their properties in suspension: TiO2 and Y2O3 to surrogate UO2 and PuO2 respectively. Once mastered with surrogate powders, this process was studied with UO2 and PuO2 to determine its optimal working conditions. The dispersion of UO2 and/or PuO2 suspensions was studied by acoustophorometry and rheology in order to optimize their formulation, as well as the freeze granulation step. Finally, UO2 and UO2-PuO2 pellets were shaped from the granules, and their sintering and microstructure were studied.

A otimização de combustíveis nucleares para uso em reatores a água pressurizada pode ser obtida pelo aumento da taxa de queima do combustível. Para isso, no entanto, é necessário levar em conta o aumento na reatividade inicial no reator, causada pelo maior enriquecimento do combustível. Esse problema, por sua vez, pode ser contornado por meio da introdução dos chamados venenos queimáveis diretamente nas pastilhas combustível de UO2. Alguns elementos do grupo das terras-raras possuem propriedades físicas e químicas que os tornam apropriados para esse uso dentro de reatores. Para caracterizar a microestrutura do combustível UO2 utilizado em reatores a água pressurizada dopado de érbio, pastilhas de UO2-Er2O3 foram preparadas, com teor de Er2O3 variando de 1,0 a 9,8 wt%, e analisadas por difração de raios X (DRX) para determinar se houve a formação de solução sólida pelo composto e determinar a variação do parâmetro de rede da solução em função da concentração de érbia. Apesar da análise por DRX ter mostrado que todo o érbio se incorporou à rede de UO2, ela também evidenciou a emergência de uma segunda fase, de estrutura do tipo fluorita, e cuja fração mássica aumenta em função do teor de érbia, enquanto seu parâmetro de rede diminui. Esses resultados são compatíveis com o fenômeno de segregação de defeitos, que consiste na formação de microdomínios segregados da rede principal nos quais há uma concentração maior dos defeitos i.e. são regiões mais ricas em érbio. Assim, a análise por DRX mostrou que houve formação de solução sólida de (U,Er)O2, mas que são necessários ajustes nos parâmetros de sinterização para que seja obtida uma solução monofásica. / Optimization of nuclear fuel for use in pressurized water reactors can be achieved by obtaining higher burnups. This, however, requires the excess reactivity caused by increasing the fuels enrichment to be taken into account, which can be done by introducing burnable absorbers into the UO2 fuel pellets themselves. Some of the rare earth elements have thermal and mechanical properties that make them appropriate for use inside the reactor. In order to characterize the microstructure of erbium-doped UO2 fuel, sintered UO2-Er2O3 pellets were prepared, with Er2O3 content ranging from 1.0 to 9.8wt%, and analyzed by X-ray diffraction to determine whether the composite formed solid solutions and, if so, evaluate the lattice parameter as a function of erbia concentration. While XRD analysis showed the Er2O3 completely dissolved in the UO2 powder, it also evidenced the emergence of a second fluorite-type phase, whose phase fraction increases and lattice parameter decreases with increasing erbia concentration. Analysis of the diffraction patterns showed this emerging phase has the same crystalline structure as the host lattice, but with a smaller lattice parameter. These results are compatible with the phenomenon of defect segregation, which consists in the formation of microdomains with a higher concentration of defects i.e. rare-earth richer regions. Thus, XRD analysis showed the formation of (U,Er)O2 solid solution, but such that there are still adjustments in the sintering parameters that need to be made in order to achieve a single-phase solid solution.

La température et la pression partielle d’oxygène (PO2) constituent les paramètres majeurs contrôlantles évolutions thermochimiques en réacteur des combustibles nucléaires de type oxyde, et notammentla spéciation des produits de fission potentiellement corrosifs (Cs, I, Te). Pour limiter les risques derupture de la gaine en Zr par corrosion, une solution innovante consiste à imposer au combustible defonctionner dans un domaine de PO2 où les espèces chimiques des gaz de fission sont inoffensives, pardopage in-situ avec un tampon oxydo-réducteur solide. Le niobium, avec ses couples redoxNbO2/NbO et Nb2O5/NbO2, a été identifié comme le candidat le plus prometteur. Un procédé defabrication d’un combustible dopé niobium répondant à cet objectif et conforme aux spécificationsd’usage (densité, microstructure) a été optimisé. L’étude expérimentale du système UO2-NbOx a révélél’existence à 810°C d’une phase liquide entre UO2 et NbO2, non identifiée à ce jour. La caractérisationdes phases solides et en solution du niobium nous a conduit à proposer un modèle thermodynamiquede solubilité du dopant dans UO2 à 1700°C. Une étude approfondie de la spéciation du niobiumprécipité a permis d’identifier la présence simultanée dans le matériau des phases majeures NbO2 etNbO, ainsi que Nb en moindre teneur. La coexistence du niobium sous deux degrés d’oxydationdifférents constitue un élément-clé de démonstration d’un possible effet tampon in-situ, dont l’impactest observé sur certaines propriétés du combustible dépendantes de la PO2, la densification notamment.Les résultats confirment le potentiel prometteur des combustibles tamponnés en PO2 au regard de sesperformances en réacteur. / Temperature and oxygen partial pressure (PO2) of nuclear oxide fuels are the main parametersgoverning both their thermochemical evolution in reactor and the speciation of volatile fissionproducts such as Cs, I or Te. An innovative way to limit the risk of cladding rupture by corrosionunder irradiation consists in buffering the oxygen partial pressure of the fuel under operation in a PO2domain where the fission gas are harmless towards Zr clad, by using solid redox buffers as additives.Niobium, with its NbO2/NbO and Nb2O5/NbO2 redox couples has been found to be a promisingcandidate to this end. A manufacturing process of a buffered UO2 fuel, doped with niobium has beenoptimized, in order to fulfill usual specifications (density, microstructure). The experimental study ofthe UO2-NbOx system has shown the existence of a liquid phase between UO2 and NbOx at 810°C,which was not reported in the literature. The characterization of Nb containing phases present in UO2both in solid solution and as precipitates has lead us to propose a solubility thermodynamic model ofniobium in UO2 at 1700°C. An extensive study of the niobium precipitates shows the co-existence inthe fuel of NbO2 and NbO as major phases, together with small amounts of metallic Nb. The coexistenceof niobium under two oxidation states inside the fuel is a key element of demonstration of apossible in-situ buffering effect, which is likely to impact some properties of the material that aredependent upon PO2, such as densification. These results confirm the promising potential of oxygenbuffered fuels as regard to their performance in reactor.

A otimização de combustíveis nucleares para uso em reatores a água pressurizada pode ser obtida pelo aumento da taxa de queima do combustível. Para isso, no entanto, é necessário levar em conta o aumento na reatividade inicial no reator, causada pelo maior enriquecimento do combustível. Esse problema, por sua vez, pode ser contornado por meio da introdução dos chamados venenos queimáveis diretamente nas pastilhas combustível de UO2. Alguns elementos do grupo das terras-raras possuem propriedades físicas e químicas que os tornam apropriados para esse uso dentro de reatores. Para caracterizar a microestrutura do combustível UO2 utilizado em reatores a água pressurizada dopado de érbio, pastilhas de UO2-Er2O3 foram preparadas, com teor de Er2O3 variando de 1,0 a 9,8 wt%, e analisadas por difração de raios X (DRX) para determinar se houve a formação de solução sólida pelo composto e determinar a variação do parâmetro de rede da solução em função da concentração de érbia. Apesar da análise por DRX ter mostrado que todo o érbio se incorporou à rede de UO2, ela também evidenciou a emergência de uma segunda fase, de estrutura do tipo fluorita, e cuja fração mássica aumenta em função do teor de érbia, enquanto seu parâmetro de rede diminui. Esses resultados são compatíveis com o fenômeno de segregação de defeitos, que consiste na formação de microdomínios segregados da rede principal nos quais há uma concentração maior dos defeitos i.e. são regiões mais ricas em érbio. Assim, a análise por DRX mostrou que houve formação de solução sólida de (U,Er)O2, mas que são necessários ajustes nos parâmetros de sinterização para que seja obtida uma solução monofásica. / Optimization of nuclear fuel for use in pressurized water reactors can be achieved by obtaining higher burnups. This, however, requires the excess reactivity caused by increasing the fuels enrichment to be taken into account, which can be done by introducing burnable absorbers into the UO2 fuel pellets themselves. Some of the rare earth elements have thermal and mechanical properties that make them appropriate for use inside the reactor. In order to characterize the microstructure of erbium-doped UO2 fuel, sintered UO2-Er2O3 pellets were prepared, with Er2O3 content ranging from 1.0 to 9.8wt%, and analyzed by X-ray diffraction to determine whether the composite formed solid solutions and, if so, evaluate the lattice parameter as a function of erbia concentration. While XRD analysis showed the Er2O3 completely dissolved in the UO2 powder, it also evidenced the emergence of a second fluorite-type phase, whose phase fraction increases and lattice parameter decreases with increasing erbia concentration. Analysis of the diffraction patterns showed this emerging phase has the same crystalline structure as the host lattice, but with a smaller lattice parameter. These results are compatible with the phenomenon of defect segregation, which consists in the formation of microdomains with a higher concentration of defects i.e. rare-earth richer regions. Thus, XRD analysis showed the formation of (U,Er)O2 solid solution, but such that there are still adjustments in the sintering parameters that need to be made in order to achieve a single-phase solid solution.

Durant l'irradiation en réacteur la fission des atomes d'actinides entraine la création de grandes quantités de défauts, qui affecte les propriétés physiques et chimiques des matériaux dans le réacteur, en particulier les matériaux combustibles ou de structure. Une des méthodes non destructives pouvant être utilisées pour caractériser les défauts induits par irradiation, vides ou contenant les produits de fission, est la spectroscopie d'annihilation de positons (SAP). Cette technique expérimentale consiste à détecter le rayonnement généré lors de l'annihilation du paire électron-positon dans un échantillon et en déduire les propriétés de la matière étudiée. Les positons peuvent être piégés dans les défauts de type lacunaire dans les solides, et en mesurant leur temps de vie et les distribution de moment du rayonnement d'annihilation, on peut obtenir des informations sur les volumes libres et les environnements chimiques des défauts. Dans ce travail, des calculs de structure électronique des caractéristiques d'annihilation de positons ont été effectués en utilisant la théorie de la fonctionnelle de la densité à deux composants (TCDFT). Pour calculer les distributions de moment rayonnement d'annihilation, nous avons implémenté les méthodes nécessaires dans le code de calcul libre ABINIT. Les résultats théoriques ont été utilités pour contribuer à l'identification des défauts d'irradiation dans deux céramiques nucléaires, le carbure de silicium (SiC) et le dioxyde d'uranium (UO2). / During in-pile irradiation the fission of actinide nuclei causes the creation of large amounts of defects, which affect the physical and chemical properties of materials inside the reactor, in particular the fuel and structural materials. Positron annihilation spectroscopy (PAS) can be used to characterize irradiation induced defects, empty or containing fission products. This non-destructive experimental technique involves detecting the radiation generated during electron-positron annihilation in a sample and deducing the properties of the material studied. As positrons get trapped in open volume defects in solids, by measuring their lifetime and momentum distributions of the annihilation radiation, one can obtain information on the open and the chemical environments of the defects. In this work electronic structure calculations of positron annihilation characteristics were performed using two-component density functional theory (TCDFT). To calculate the momentum distributions of the annihilation radiation, we implemented the necessary methods in the open-source ABINIT program. The theoretical results have been used to contribute to the identification of the vacancy defects in two nuclear ceramics, silicon carbide (SiC) and uranium dioxide (UO2).

La reproduction expérimentale de transitoires thermiques accidentels de type Accident par Perte de Réfrigérant Primaire (APRP) en laboratoire a permis d’observer la fragmentation du combustible fortement irradié lorsque la gaine se déforme sous l’augmentation de la température. Ces fragments de petites tailles peuvent se relocaliser dans le ballon voire être éjectés hors du crayon cas de rupture de gaine. La zone High Burnup Structure (HBS) des combustibles fortement irradiés est la plus susceptible de se fragmenter et d’être relocalisée par sa position en périphérie de pastille. Pour expliquer ce phénomène, l’hypothèse retenue est que le transitoire provoque une surpression dans les bulles HBS ce qui mène à la décohésion des joints de grains et à la fragmentation. Cette thèse a pour but de développer un critère de fissuration mécanique du combustible pour mieux comprendre le comportement des bulles HBS lors des conditions thermiques APRP. Ce travail se base sur une méthode une méthode micromécanique en trois étapes : i) la représentation qui permet de caractériser la microstructure de la zone HBS (leurs dimensions, leur fraction volumique, et la pression interne). Deux sources d’informations seront utilisées : les observations expérimentales provenant de disques ou de pastilles de combustible irradiés à fort taux de combustion et d’outils numériques(avec Alcyone-Caracas [JSB+14]) / Under Loss Of Coolant Accident(LOCA) transients conditions, the high irradiated fuel is fragmented in small sizes fragments who can be relocated in the balloon, or being ejected out of the fuel rod if the latter burst. This work focuses on the pellet rim, where bubbles density increases owing to a higher irradiation level. Usually the hypothesis used to explain fuel fragmentation during transient is grain cleavage induced by over pressurized fission gas bubbles, located at the grain boundary. The aim of this study is to define a macroscopic fragmentation model based on a micro mechanical approach to have a better understanding of the fuel mechanical behaviour at lower scale : size and volume fraction of fragments. This PhD introduces a stepwise micromechanical method based on three steps : i) firstly, we detail how to model the HBS microstructure including pressurized porosities, based on experimental or numerical data and define a representative volume element (RVE)

No description available.

L'irradiation d'un solide par un faisceau d'ions peut conduire à l'émission d'atomes, de molécules ou d'agrégats, neutres ou ionisés, traduisant la mise en mouvement des atomes à proximité de la surface. L'étude de la pulvérisation permet alors de mieux comprendre l'état de la matière résultant de l'excitation induite par le passage de l'ion. Dans le cas d'ions lents multichargés, la neutralisation du projectile, au dessus de la surface, peut conduire à une forte excitation électronique dans un domaine de vitesse où les collisions élastiques avec les atomes de la cible sont à l'origine du ralentissement de l'ion incident. L'etude de l'effet de la charge initiale de tels ions sur le processus de pulvérisation, a été réalisée par la mesure absolue des distributions angulaires d'émission d'atomes d'uranium depuis une surface de dioxyde d'uranium. Les expériences ont été réalisées en deux étapes, la collection des particules émises sur un substrat durant l'irradiation étant suivie d'une analyse, par Spectroscopie de Rétrodiffusion Rutherford (RBS), de la surface de ces derniers. Cette methode permet de caractériser l'émission des particules neutres, qui représentent la vaste majoritée des éjecta. Les résultats obtenus donnent accès à l'évolution du processus de pulvérisation, en fonction de l'état de charge d'ions lourds xénon, sur une gamme d'énergie cinétique s'étendant de 1,5 à 81 keV et pour différents angles d'incidence des projectiles. Il apparaît, entre autres, un accroissement important du rendement de pulvérisation avec l'état de charge initial sous incidence normale.

<p>The kinetics for radiation induced dissolution of spent nuclear fuel is a key issue in the safety assessment of a future deep repository. Spent nuclear fuel mainly consists of UO₂ and therefore the release of radionuclides (fission products and actinides) is assumed to be governed by the oxidation and subsequent dissolution of the UO₂ matrix. The process is influenced by the dose rate in the surrounding groundwater (a function of fuel age and burn up) and on the groundwater composition. In this licentiate thesis the effects of HCO₃- (a strong complexing agent for UO2²⁺) and ionic strength on the kinetics of UO₂ oxidation and dissolution of oxidized UO₂ have been studied experimentally.</p><p>The experiments were performed using aqueous UO₂particle suspensions where the oxidant concentration was monitored as a function of reaction time. These reaction systems frequently display first order kinetics. Second order rate constants were obtained by varying the solid UO₂surface area to solution volume ratio and plotting the resulting pseudo first order rate constants against the surface area to solution volume ratio. The oxidants used were H₂O₂(the most important oxidant under deep repository conditions), MnO₄- and IrCl₆^2-. The kinetics was studied as a function of HCO₃- concentration and ionic strength (using NaCl and Na₂SO₄as electrolytes).</p><p>The rate constant for the reaction between H₂O₂ and UO₂ was found to increase linearly with the HCO3- concentration in the range 0-1 mM. Above 1 mM the rate constant is independent of the HCO3- concentration. The HCO₃- concentration independent rate constant is interpreted as being the true rate constant for oxidation of UO₂ by H₂O₂ [(4.4 ± 0.3) x 10-6 m min-1] while the HCO3- concentration dependent rate constant is used to estimate the rate constant for HCO₃- facilitated dissolution of UO₂2+ (oxidized UO₂) [(8.8 ± 0.5) x 10-3 m min-1]. From experiments performed in suspensions free from HCO₃- the rate constant for dissolution of UO₂2+ was also determined [(7 ± 1) x 10^-8mol m^-2 s^-1]. These rate constants are of significant importance for simulation of spent nuclear fuel dissolution.</p><p>The rate constant for the oxidation of UO₂ by H₂O₂ (the HCO₃- concentration independent rate constant) was found to be independent of ionic strength. However, the rate constant for dissolution of oxidized UO₂ displayed ionic strength dependence, namely it increases with increasing ionic strength.</p><p>The HCO₃- concentration and ionic strength dependence for the anionic oxidants is more complex since also the electron transfer process is expected to be ionic strength dependent. Furthermore, the kinetics for the anionic oxidants is more pH sensitive. For both MnO₄- and IrCl₆2- the rate constant for the reaction with UO₂was found to be diffusion controlled at higher HCO3- concentrations (~0.2 M). Both oxidants also displayed ionic strength dependence even though the HCO₃- independent reaction could not be studied exclusively.</p><p>Based on changes in reaction order from first to zeroth order kinetics (which occurs when the UO₂ surface is completely oxidized) in HCO₃- deficient systems the oxidation site density of the UO₂ powder was determined. H₂O₂and IrCl₆2- were used in these experiments giving similar results [(2.1 ± 0.1) x 10-4 and (2.7 ± 0.5) x 10^-4 mol m^-2, respectively].</p>

<p>Denna licensiatavhandling behandlar oxidativ upplösning av UO2. Upplösning av UO2 studeras huvudsakligen då UO2-matrisen hos använt kärnbränsle förväntas fungera som en barriär mot frigörande av radionuklider i ett framtida djupförvar. Lösligheten av U(IV) är mycket låg under i djupförvaret rådande förhållanden emedan U(VI) har betydligt högre löslighet. Oxidation av UO2-matrisen kommer därför att påverka dess löslighet och därmed dess funktion som barriär. I denna avhandling studeras den relativa effektiviteten av en- och två-elektronoxidanter för upplösning av UO2. Vid låga oxidantkoncentrationer är utbytet för upplösningen för en-elektronoxidanter signifikant lägre än för två-elektronoxidanter. För en-elektronoxidanter ökar dock utbytet med ökande oxidanthalt, vilket kan förklaras av den ökade sannolikheten för två konsekutiva en-elektronoxidationer av samma reaktionssite och den ökade möjligheten till disproportionering.</p><p>Radikaler och molekylära radiolysprodukters relativa inverkan på oxidativ upplösning av UO2 studeras också i denna avhandling genom mätning av mängden upplöst U(VI) i γ-bestrålade system som dominerades av olika oxidanter. Dessa studier visade att upplösningshastigheten av UO2 kan uppskattas från oxidantkoncentrationer framtagna genom simuleringar av radiolys i motsvarande homogena system och hastighetskonstanterna för ytreaktionerna. Simuleringarna visar att de molekylära oxidanterna kommer vara de viktigaste oxidanterna i alla system i denna studie vid långa bestrålningstider (>10 timmar). Vid liknande simuleringar av α-bestrålade system fanns att vid förhållanden relevanta för ett djupförvar för använt kärnbränsle, är det endast de molekylära oxidanterna (i huvudsak H2O2) som är av betydelse för upplösningen av bränslematrisen.</p><p>Då använt kärnbränsle innehåller en mängd radionuklider som utsätter UO2-matrisen för kontinuerlig bestrålning, är det av vikt att undersöka hur bestrålning påverkar reaktiviteten av UO2. Bestrålningseffekten på reaktionen mellan UO2 och MnO4- studerades. Dessa försök visade att bestrålning av UO2 vid doser >40 kGy leder till att reaktiviteten ökar upp till 1.3 gånger reaktiviteten av obestrålad UO2. Den ökade reaktiviteten kvarstår efter bestrålningen och effekten kan därför möjligen tillskrivas permanenta förändringar i materialet. Vid uppskattning av reaktiviteten hos använt kärnbränsle måste hänsyn tas till denna effekt då bränslet redan efter ett par dagar i reaktor blivit utsatt för doser >40 kGy.</p><p>Det har tidigare föreslagits att hastigheten för en heterogen västka/fast-fas reaktion är beroende av partikelstorleken hos det fasta materialet, vilket har studerats för UO2-partiklar i denna avhandling. Experimentellt bestämda kinetiska parametrar jämförs med de föreslagna ekvationerna för fyra storleksfraktioner av UO2-pulver och en UO2-pellet. Studien visade partikelstorleksberoendet av andra ordningens hastighetskonstant och aktiveringsenergin för oxidation av UO2 med MnO4- beskrivs relativt väl av de föreslagna ekvationerna.</p> / <p>The general subject of this thesis is oxidative dissolution of UO₂. The dissolution of UO₂ is mainly investigated because of the importance of the UO₂ matrix of spent nuclear fuel as a barrier against radionuclide release in a future deep repository. U(IV) is extremely insoluble under the reducing conditions prevalent in a deep repository, whereas U(VI) is more soluble. Hence, oxidation of the UO₂-matrix will affect its solubility and thereby its function as a barrier. In this thesis the relative efficiency of one- and two electron oxidants in dissolving UO₂is studied. The oxidative dissolution yield of UO₂was found to differ between one- and two-electron oxidants. At low oxidant concentrations the dissolution yields for one-electron oxidants are significantly lower than for two-electron oxidants. However, the dissolution yield for one-electron oxidants increases with increasing oxidant concentration, which could be rationalized by the increased probability for two consecutive one-electron oxidations at the same site and the increased possibility for disproportionation.</p><p>Furthermore, the relative impact of radical and molecular radiolysis products on oxidative dissolution of UO₂is investigated. Experiments were performed where the amount of dissolved U(VI) was measured in γ-irradiated systems dominated by different oxidants. We have found that the UO₂dissolution rate in systems exposed to γ-irradiation can be estimated from oxidant concentrations derived from simulations of radiolysis in the corresponding homogeneous systems and rate constants for the surface reactions. These simulations show that for all systems studied in this work, the molecular oxidants will be the most important oxidants for long irradiation times (>10 hours). Similar simulations of α-irradiated systems show that in systems relevant for a deep repository for spent nuclear fuel, only the molecular oxidants (mainly H₂O₂) are of importance for the dissolution of the fuel matrix.</p><p>The effect on UO₂ reactivity by irradiation of the material is of importance when predicting the spent fuel dissolution rate since the fuel, due to its content of radionuclides, is exposed to continuous self-irradiation. The effect of irradiation on the reaction between solid UO₂and MnO₄^- in aqueous solutions was studied. It was found that irradiation of UO2 at doses >40 kGy increases the reactivity of the material up to ~1.3 times the reactivity of unirradiated UO₂. The increased reactivity remains after the irradiation and can possibly be attributed to permanent changes in the material. This issue must be taken into account when predicting the reactivity of spent nuclear fuel since the fuel is exposed to doses >40 kGy after only a few days in the reactor.</p><p>It has earlier been suggested that the rate of a heterogeneous liquid-solid reaction depends on the size of the solid particles. This was investigated for UO₂particles in this thesis. Experimental kinetic parameters are compared to the previously proposed equations for UO₂ powder of four size fractions and a UO₂ pellet. We have found that the particle size dependence of the second order rate constant and activation energy for oxidation of UO₂ by MnO₄^- is described quite well by the proposed equations.</p>