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. 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. 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. 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. / 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 UO2 is studied. The oxidative dissolution yield of UO2 was 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. Furthermore, the relative impact of radical and molecular radiolysis products on oxidative dissolution of UO2 is 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 UO2 dissolution 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. 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 UO2 and 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. 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 UO2 particles 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. / QC 20101123

Le combustible nucléaire usé, constitué d’uranium, de plutonium et de produits de fission (palladium, technétium, molybdène etc.), est traité industriellement en France par le procédé PUREX. Ce procédé hydrométallurgique permet d’extraire l’uranium et le plutonium à l’aide d’une phase organique composée de tri-n-butylphosphate (TBP) dilué dans un mélange d’hydrocarbures aliphatiques (TPH). Au cours des différents cycles d’extraction, la phase organique est soumise à des phénomènes d’hydrolyse acide et de radiolyse conduisant à la formation de produits de dégradation. Afin d’assurer le recyclage du solvant organique et le bon fonctionnement du procédé industriel, un traitement approprié permet d’éliminer ces produits. Néanmoins, après plusieurs décennies de fonctionnement des usines, des encrassements de certains équipements ont été observés. Ces solides contiennent les éléments palladium, carbone, oxygène et azote. Industriellement, une solution curative a été mise en œuvre pour éliminer ces crasses. Il convient toutefois de comprendre les phénomènes qui ont conduit à leur formation. Ce travail de doctorat a pour objectif d’étudier le comportement physico-chimique du palladium dans les cycles d’extraction et les mécanismes conduisant à la formation de solides. Pour cela, un système biphasique, constitué d’une phase organique de TBP dilué dans le TPH et d’une phase aqueuse d’acide nitrique contenant du nitrate de palladium(II), a été irradié par une source externe de rayonnement gamma afin de simuler le vieillissement du solvant d’extraction. Des précipités non radioactifs contenant du palladium et représentatifs des encrassements industriels ont été obtenus. Des analyses multi-techniques de ces précipités ont montré que les solides sont constitués d’un mélange complexe de cyanure de palladium(II), de carboxylate de palladium(II) et de divers produits organiques (di-n-butylphosphate (HDBP), composés à fonctions amines et hydrazine). Les produits menant à la précipitation du palladium proviennent de la dégradation du TBP et de celle du TPH et sont présents en phase aqueuse et en phase organique.Certains produits de dégradation formés lors de l’irradiation semblent favoriser l’extraction du palladium. Pour comprendre le rôle des produits de dégradation sur l’extraction du palladium, le solvant d’extraction utilisé dans le procédé PUREX a été dopé en composés organiques. Les composés testés sont des composés commerciaux représentatifs de produits de dégradation identifiés un solvant dégradé. De plus, les espèces testées sont des composés susceptibles d’interagir avec le palladium(II) (cétone, acide carboxylique, alcool, aldéhyde et alcène). Parmi ces composés, seul le 5-dodécène conduit à une augmentation significative de l’extraction du palladium(II) par rapport à une phase organique non dopée et à la formation d’un précipité noir constitué de palladium métal. Les analyses réalisées sur la phase organique dopée en 5-dodécène ont mis en évidence la formation d’un complexe ternaire dans lequel le palladium serait entouré de molécules de TBP, de 5-dodécène et d’ions nitrate. Par ailleurs, une fraction du palladium(II) extrait dans la phase organique oxyde le 5-dodécène en cétone et est réduit en palladium métal (précipité noir). Des voies de formation des différents composés constituant les crasses ont été proposées. Les alcènes, les cétones et l’acide nitreux apparaissent comme des précurseurs dans la formation de cyanure de palladium(II). Un carboxylate de palladium(II) serait formé par réaction de palladium(II) avec l’acide carboxylique correspondant. Les amines primaires pourraient se former par hydrogénation des nitroalcanes catalysé par la présence de palladium métal tandis que l’hydrazine pourrait se former par radiolyse des amines primaires. / The spent nuclear fuel composed of uranium, plutonium and fission products (such as palladium, technetium, molybdenum, etc.) is industrially processed in France by the PUREX process. This hydrometallurgical process allows separating uranium and plutonium by using an organic phase composed of tri-n-butyl phosphate (TBP) diluted in a mixture of aliphatic hydrocarbons (TPH = tetra propylene hydrogen). During the various extraction cycles, the organic phase is subjected to acid hydrolysis and radiolysis phenomena leading to the formation of degradation products. To ensure the recycling of the organic solvent and a normal operation of the industrial process, an appropriate treatment eliminates these degradation products. Nevertheless, after several decades of industrial operation of the factories, precipitates have been observed in some equipments. These solids contain palladium, carbon, oxygen and nitrogen atoms. Industrially, a curative solution has been found and allows eliminating these cruds containing in particular palladium. However, it is important to understand the formation of palladium precipitates in the liquid-liquid extraction cycles. This thesis aims to study the behavior of palladium in the extraction cycles and the mechanisms leading to the formation of solids.For this purpose, a biphasic system containing an organic phase of TBP diluted in TPH and a aqueous phase of nitric acid containing palladium(II) nitrate has been irradiated by an external source (gamma-irradiation). This irradiation allows simulating the aging of the extraction solvent. Nonradioactive precipitates which are representative of the industrial crud were obtained. A multi-technical analysis of these inactive precipitates has shown that the solids are composed of a complex mixture: (i) palladium cyanide(II), (ii) palladium(II) carboxylate and (iii) various organic products (di-n-butyl phosphate (HDBP)), compounds with amine functions and hydrazine. Moreover, we demonstrate that the degradation products leading to the precipitation of palladium come from either TBP or TPH and are present in the aqueous phase and in the organic phase.Some degradation products formed during irradiation seem to favor the extraction of palladium(II). To understand the role of degradation products on palladium extraction, the extraction solvent used in the PUREX process was doped with organic compounds. The compounds tested are commercial compounds representative of degradation products identified a degraded solvent. In addition, the compounds tested are species which can react with palladium(II) (ketone, carboxylic acid, alcohol, aldehyde and alkene). Among these compounds, only 5-dodecene leads to a significant increase in the extraction of palladium(II) in comparison with an undoped organic phase and the formation of a black precipitate composed of palladium metal. The analyses carried out on the 5-dodecene-doped organic phase revealed the formation of a ternary complex in which the palladium would be surrounded by TBP molécules, 5-dodecene and nitrate ions. In addition, a fraction of the palladium(II) extracted in the organic phase oxidizes the 5-dodecene to ketone and is reduced to palladium metal (black precipitate).Formation pathways of the various compounds present in the nonradioactive solids have been proposed. Alkenes, ketones and nitrous acid appear as precursors in the formation of palladium cyanide(II). A palladium(II) carboxylate would be formed by reaction of palladium(II) with the corresponding carboxylic acid. Primary amines could be formed by hydrogenation of nitroalkanes catalyzed by the presence of palladium metal while hydrazine could be formed by radiolysis of primary amines.Key word: palladium, radiolysis, PUREX process, degradation products, precipitates

Afin de conditionner les déchets technologiques issus du démantèlement de l’Atelier de Vitrification de Marcoule au Commissariat à l’Energie Atomique et aux énergies alternatives (CEA), leur blocage dans une matrice cimentaire est envisagé. Dans ce contexte, l’effet des rayonnements ionisants issus des déchets nucléaires sur la matrice de confinement doit être examiné afin de garantir d’une part l’intégrité du colis, et d’autre part une production de gaz de radiolyse limitée. Ce travail de thèse s’intéresse au comportement sous irradiation gamma et électronique de trois types de matériaux cimentaires aux constituants différents et se focalise sur la production de gaz de radiolyse et sur l’évaluation de la résistance physique à travers l’observation des modifications structurales. Le sujet est traité par une double approche à la fois sur pâte de ciment et sur phases modèles, c’est-à-dire sur les hydrates constitutifs des pâtes de ciment synthétisés indépendamment. Il apparaît clairement que l’eau porale ainsi que les hydrates constitutifs contribuent à la production d’hydrogène radiolytique, avec une forte variation selon la nature des matériaux cimentaires. Ainsi, les ciments à base d’aluminates de calcium et les ciments phospho-magnésiens présentent un intérêt notable par rapport aux ciments silico-calciques usuels quant à la production d’hydrogène. Aux très fortes doses (plusieurs GGy) la résistance structurale sous irradiation électronique a été évaluée par diffraction de rayons X. Les hydrates constitutifs des trois familles de ciment étudiées présentent une bonne résistance structurale. Malgré la présence de variations dimensionnelles et microstructurales, ils ne s’amorphisent pas sous irradiation, ce qui s’avère positif en vue de l’application industrielle envisagée. / In order to treat the technological waste arising from the dismantling of the Marcoule Vitrification facility of the French Atomic Energy Commission (CEA), conditioning in a cement matrix is being put forward. Within this context, the impact of ionizing radiation produced by the nuclear waste on the confinement matrix ought to be investigated in order to ensure both the integrity of the package and the limitation of the radiolytic gas production. This thesis investigates the behavior of three types of cement compounds with distinct constituents under gamma and electronic radiation. This study deals with both the radiolytic gas production and the physical resistance of the materials using a structural modification examination. A double and complementary approach is used treating cement pastes and synthetic cement compounds together. It clearly appears that the pore water and the hydrates themselves both contribute to the radiolytic hydrogen production, with a significant variation depending on the nature of the materials. As far as radiolysis is concerned, calcium aluminate-based cements and magnesium phosphate cements are of considerable interest in comparison with the usual calcium silicate cements. At very high doses (GGy range), the structural resistance under electron irradiation was evaluated by X-ray diffraction. The constituent hydrates of the three cement types studied exhibit a good structural resistance. Despite the presence of dimensional variations at the unit cell scale as well as microstructural evolution, no amorphization is observed under irradiation, which is an interesting result with respect to the intended industrial application.

Alkaline fuel cells are a promising technology, with their sturdy design and many applications they are held back mostly by their cost. By introducing a catalyst, the activation energy of the cell can be reduced to an overcomable amount. Unfortunately, due to the high cost and sparse availability of the most used catalyst metal today, platinum, it has become apparent that a new suitable catalyst must be found in order to make the fuel cells economically feasible. Silver and palladium have been proposed as promising alternatives, sharing a majority of the traits but with a fraction of the cost. The original aim of this project was to study the performance of electrodes in an alkaline electrolyte loaded with different ratios of palladium and silver. However, due to the COVID-19 situation the project was not able to be completed and the aim of the project changed. The new aim was divided into two parts. The first one being to study how the initial concentration of silver ions affects the size of the obtained particles. This was achieved by a radiolysis-based method of synthesis in an aqueous solution. The second aim was to study the performance of the electrodes loaded with different amounts of silver and different average particle size. However, this part was not possible to conduct either. Therefore, results from a previous study performed by I. L. Soroka et al. was used for discussion. The results point towards a lower initial concentration achieving a smaller average particle size and a lower loading of catalyst on the electrode can be compensated by a smaller average particle size of the catalyst.

The core of a BWR constitutes a highly complex radiational and chemical environment. Nuclear fission is utilized to generate power and as a consequence, large quantities of ionizing radiation are produced. Gamma and beta-particles along with neutrons have the sufficient ranges to escape the fuel rods and deposit energy in the reactor coolant. By doing so, radiolysis is initiated. The radiolysis species present in the coolant can interact chemically with the fuel rod cladding and cause corrosion. Different forms of corrosion are found on BWR fuel, with some effects being very local.  This thesis work outlines a method developed to investigate local corrosion phenomena, with a high spatial resolution. The purpose was to study Zircaloy corrosion and more particularly, to investigate an observed jump in corrosion thickness around the lower enrichment step on BWR fuel rod cladding. The corrosion thickness jump is a very local effect, hence the need for high spatial resolution. Monte-Carlo simulations were performed in a GEANT4-model of an inoperation reactor, to study the energy deposited from ionizing radiation in the coolant, around the low enrichment-step of the fuel rods. The energy dose data was then used as input to lwrChem to compute electrochemical potential and equilibrium concentrations of radiolysis species. These are the quantities needed to compute an equivalent corrosion rate on the cladding surface, although this was not performed within this project. The main focus was to successfully develop the two-program method using GEANT4 and lwrChem and this was achieved.  The project was performed Uppsala University with financial contribution from Vattenfall Nuclear Fuel AB and scientific contribution from Studsvik Nuclear AB.

No description available.

Water is a major component of living organisms, which can be 70-85% of the weight of cells. For this reason, water is a main target of ionizing radiations and plays a central role in radiobiology. Heavy ions, electrons and photons interact with water molecules; mainly by ionization and excitation. Neutrons interact with water molecules by elastic interactions, which generate recoil ions that will create ionizations and excitations in water molecules. These fast events (~10[superscript -12] s) lead to the formation of Reactive Oxygen Species (ROS). The ROS, in particular the hydroxyl radical (¨OH), interact with neighbour molecules such as proteins, lipids and nucleic acids by chemical interaction. Microbeams can irradiate selectively either the external membrane, the cytoplasm and the cell nucleus. These studies have shown that cell survival is greatly reduced when the nucleus is irradiated, but that this is not the case when cytoplasm or cell membrane is irradiated. Thus, DNA is a very sensitive site to ionizing radiation and ROS. For this reason, DNA has long been considered the most important molecule to explain radiobiological effects such as cell death. However, this concept has been challenged recently by new experimental results that have shown that cells which have not been directly in contact with radiation are also affected. This is called the bystander effect. Further studies have shown that a group of cells and their environment reacts collectively to radiation. A hypothesis put forward to explain this radiobiological phenomenon is that a irradiated cell will secrete signalling molecules that will affect non-irradiated cells. The implicated phenomenon and molecules are poorly understood at this moment. The purpose of this work is to improve our comprehension of the phenomenon in the microsecond that follows the irradiation. To these ends, a new Monte-Carlo simulation program of water radiolysis by photons has been generated. For photons of energy <2 MeV, they interact with water mainly by Compton and photoelectric effects, which create energetic electrons in water. The created electrons are then followed by our existing programs to simulate the radiolysis of water by photons. Similarly, a new code has been built to simulate the neutrons interaction with water. This code simulates the elastic collisions of a neutron with water molecules and calculates the number and energy of recoil protons and oxygen ions. The main part of this Ph.D. work was the generation of a non-homogeneous Monte-Carlo Step-By-Step (SBS) simulation code of non-homogeneous radiation chemistry. This new program has been used successfully to simulate radiolysis of water by ions of various LET, pH, ion types ([superscript 1]H[superscript +], [superscript 4]He[superscript 2+], [superscript 12]C[superscript 6+]) and temperature. The program has also been used to simulate the dose-rate effect and the Fricke and Ceric dosimeters. More complex systems (glycine, polymer gels and HCN) have also been simulated.

Dissolution of the UO2 matrix is one of the potential routes for radionuclide release in a future deep geological repository for spent nuclear fuel. This doctoral thesis focuses on interfacial reactions of relevance in radiation-induced dissolution of UO2 and is divided in two parts: In the first part, we sought to explore the effects of solid phase composition: The impact of surface stoichiometry on the reactivity of UO2 towards aqueous radiolytic oxidants was studied. H2O2 reacts substantially faster with stoichiometric UO2 than with hyperstoichiometric UO2. In addition, the release of uranium from stoichiometric UO2 is lower than from hyperstoichiometric UO2. The behavior of stoichiometric powder changes with exposure to H2O2, approaching the behavior of hyperstoichiometric UO2 with the number of consecutive H2O2 additions. The impact of Gd-doping on the oxidative dissolution of UO2 in an aqueous system was investigated. A significant decrease in uranium dissolution and higher stability towards H2O2 for (U,Gd)O2 pellets compared to standard UO2 was found. In the second part, we sought to look at the effect of external factors: The surface reactivity of H2 and O2 was studied to understand the overall oxide surface reactivity of aqueous molecular radiolysis products. The results showed that hydrogen-abstracting radicals and H2O2 are formed in these systems. Identical experiments performed in aqueous systems containing UO2 powder showed that the simultaneous presence of H2 and O2 enhances the oxidative dissolution of UO2 compared to a system not containing H2. The effect of groundwater components such as bentonite and sulfide on the oxidative dissolution of UO2 was also explored. The presence of bentonite and sulfide in water could either delay or prevent in part the release of uranium to the environment. The Pd catalyzed H2 effect is more powerful than the sulfide effect. The poisoning of Pd catalyst is not observed under the conditions studied. / <p>QC 20170421</p>

Les oxydes d’actinides sont des matériaux hygroscopiques. L’adsorption de l’eau à leurs surfaces est susceptible d’entrainer des modifications quant à la nature ou à son état. Dans le cas des oxydes à fort débit de dose, vient s’ajouter les effets de radiolyse de l’eau, entrainant sa décomposition et générant du dihydrogène. Ces deux aspects, étude de surface et radiolyse de l’eau, ont été étudiés ici. L’étude de la génération de dihydrogène par radiolyse de l’eau adsorbée à la surface a montré que cette génération linéaire dans les premiers temps atteint une concentration stable au bout de plusieurs heures. Cet état stationnaire a été très peu observé, et est absent dans le cas d’humidité importante. Un conditionnement dans une atmosphère contenant du dihydrogène a permis de mettre en lumière une réaction de consommation du dihydrogène par le matériau. Ces expériences ont permis de faire émerger un modèle cinétique basé sur deux réactions de production et de consommation de dihydrogène. La première correspond à la décomposition de l’eau sous l’effet du rayonnement, et pour la seconde il est suspecté une réduction partielle de la surface avec la formation d’une phase sous-stœchiométrique en surface. Cependant, aucune technique d’analyse de surface n’a permis de mettre en évidence formellement cette phase. La chromatographie gazeuse inverse est une technique peu intrusive vis-à-vis des couches d’eau adsorbée du fait des températures et des pressions mis en jeux et de l’absence de dépôt d’énergie. Cette technique a été employée sur des oxydes de thorium et d’uranium. Sur oxyde de thorium, il en résulte un impact de la température de calcination, avec un maximum d’énergie de surface pour une calcination à 650°C. Par ailleurs, il a été montré que la préparation du dioxyde de thorium pouvait impacter l’état de sa surface. En effet, il a été observé une déshydratation de l’oxalate de thorium au fil du temps, impactant la structure de ce dernier. Cette modification se répercute sur la surface de l’oxyde final par une chute de l’énergie de surface et par une modification sur la répartition des sites d’adsorption en surface. Néanmoins un traitement chimique de l’oxalate permet de retrouver la réactivité de surface et une distribution des sites d’adsorption. L’hydratation de la surface montre une augmentation de l’énergie de surface, mais cette augmentation n’est observée que pour des hydratations de longues durées. / Actinide oxides are hygroscopic materials. The adsorption of water on their surfaces is likely to cause changes in the nature or condition. In the case of oxides with a high dose rate, the effects of radiolysis of the water causes the decomposition of water and generates hydrogen. These two aspects: surface study and radiolysis of water have been studied here.The study of the generation of dihydrogen by radiolysis of water adsorbed on the surface has shown that this linear generation in the early stages reaches a stable concentration after several hours. This stationary state has been very little observed, and is absent in the case of significant humidity. Conditioning in a dihydrogen-containing atmosphere made it possible to highlight a reaction of consumption of dihydrogen by the material. These experiments led to the emergence of a kinetic model based on two reactions of production and consumption of dihydrogen. The first corresponds to the decomposition of the water under the effect of the radiation, and for the second it is suspected a partial reduction of the surface with the formation of a sub-stoichiometric phase on the surface, however no technique of analysis of surface has not formally highlighted this phase.Inverse gas chromatography is a technique that is not very intrusive with respect to the adsorbed water layers because of the temperatures and pressures involved and the absence of energy deposition. This technique has been used on oxides of thorium and uranium. On thorium oxide, this results in an impact of the calcination temperature, with a maximum of surface energy for calcination at 650 ° C. In addition, it has been shown that the preparation of thorium dioxide can impact the state of its surface. Indeed, it has been observed dehydration of thorium oxalate over time, impacting the structure of the latter. This modification affects the surface of the final oxide by a drop-in surface energy and a change in the distribution of surface adsorption sites. Nevertheless, a chemical treatment of oxalate makes it possible to recover the surface reactivity and a distribution of the adsorption sites. The hydration of the surface shows an increase in surface energy, but this increase is observed only for hydrations of long duration.Keywords: water sorption, radiolysis, plutonium, inverse gas chromatography, thorium, uranium

Three models have been developed and applied in the performance assessment of a final repository. They are based on accepted theories and experimental results for known and possible mechanisms that may dominate in the oxidative dissolution of spent fuel and the release of nuclides from a canister. Assuming that the canister is breached at an early stage after disposal, the three models describe three sub-systems in the near field of the repository, in which the governing processes and mechanisms are quite different. In the model for the oxidative dissolution of the fuel matrix, a set of kinetic descriptions is provided that describes the oxidative dissolution of the fuel matrix and the release of the embedded nuclides. In particular, the effect of autocatalytic reduction of hexavalent uranium by dissolved H2, using UO2 (s) on the fuel pellets as a catalyst, is taken into account. The simulation results suggest that most of the radiolytic oxidants will be consumed by the oxidation of the fuel matrix, and that much less will be depleted by dissolved ferrous iron. Most of the radiolytically produced hexavalent uranium will be reduced by the autocatalytic reaction with H2 on the fuel surface. It will reprecipitate as UO2 (s) on the fuel surface, and thus very little net oxidation of the fuel will take place. In the reactive transport model, the interactions of multiple processes within a defective canister are described, in which numerous redox reactions take place as multiple species diffuse. The effect of corrosion of the cast iron insert of the canister and the reduction of dissolved hexavalent uranium by ferrous iron sorbed onto iron corrosion products and by dissolved H2 are particularly included. Scoping calculations suggest that corrosion of the iron insert will occur primarily under anaerobic conditions. The escaping oxidants from the fuel rods will migrate toward the iron insert. Much of these oxidants will, however, be consumed by ferrous iron that comes from the corrosion of iron. The nonscavenged hexavalent uranium will be reduced by ferrous iron sorbed onto the iron corrosion products and by dissolved hydrogen. In the transport resistance network model, the transport of reactive actinides in the near field is simulated. The model describes the transport resistance in terms of coupled resistors by a coarse compartmentalisation of the repository, based on the concept that various ligands first come into the canister and then diffuse out to the surroundings in the form of nuclide complexes. The simulation results suggest that carbonate accelerates the oxidative dissolution of the fuel matrix by stabilizing uranyl ions, and that phosphate and silicate tend to limit the dissolution by the formation of insoluble secondary phases. The three models provide powerful tools to evaluate "what if" situations and alternative scenarios involving various interpretations of the repository system. They can be used to predict the rate of release of actinides from the fuel, to test alternative hypotheses and to study the response of the system to various parameters and conditions imposed upon it. / QC 20100521