Dans un contexte d'accroissement des marges de sûreté des réacteurs de générations II et III vis-à-vis des scénarios accidentels, des efforts importants de recherche sont consacrés à l'amélioration de la microstructure du combustible MOX à l'issue de son procédé de fabrication. Les deux caractéristiques microstructurales recherchées sont l'accroissement de l'homogénéité de répartition du plutonium et l'augmentation de la taille de grain. Dans cette optique, une solution envisagée est l'ajout lors du procédé de fabrication et sans modification de celui-ci, de sesquioxyde de chrome Cr2O3. Une précédente thèse sur le sujet a permis de proposer un modèle d'homogénéisation de la répartition du plutonium suite à l'ajout de Cr2O3. L'auteur a souligné l'importance de la formation du précipité PuCrO3 aux joints de grains lorsque la solubilité du chrome dans la matrice (U,Pu)O2 est atteinte. Cependant, les mécanismes d'action du chrome n'ont été étudiés que pour une atmosphère de frittage unique. Plusieurs points restent également à approfondir, notamment la solubilité du chrome et les conditions optimales de formation du précipité PuCrO3. Dans un premier temps, une étude de la spéciation du chrome solubilisé et précipité dans l'oxyde mixte (U,Pu)O2 a été réalisée. Les techniques ayant permis d'analyser directement le chrome sont la microsonde électronique et la spectroscopie d'absorption des rayons X. Il a été montré que le degré d'oxydation et l'environnement du chrome solubilisé sont indépendants de la pression partielle d'oxygène imposée lors du frittage et de la teneur en plutonium de l'oxyde mixte. La nature des précipités et la solubilité du chrome dépendent, quant à eux, de la variable thermodynamique et de la teneur en Pu. Sur la base de ces résultats, un modèle de solubilité du chrome dans l'oxyde mixte (U,Pu)O2-x a été construit. Ce modèle a été réalisé en fonction de la teneur en plutonium y de la solution solide (U1-yPuy)O2-x (y = 0,11 ; 0,275 et 1) et sur la gamme de potentiel d'oxygène d'intérêt pour le frittage du combustible (-445 kJ.mol -1< µO2 < -360 kJ.mol -1). Outre l'optimisation du dopage, ce modèle permet de définir les conditions optimales de formation du précipité PuCrO3 en fonction de la teneur en plutonium et de l'atmosphère de frittage. Dans un second temps, nous avons regardé si les conditions d'obtention du précipité PuCrO3 correspondaient à un accroissement de l'homogénéité de répartition du plutonium et une taille de grains maximale. Pour ce faire, des échantillons fabriqués avec ou sans présence de chrome et frittés sous différentes atmosphères ont été étudiés. Il a été mis en évidence que la cinétique d'interdiffusion U-Pu est complètement modifiée en présence de chrome. De plus, suite à l'ajout de chrome, les conditions permettant d'accroître la cinétique d'interdiffusion U-Pu ne sont pas forcément associées à une taille de grain maximale.A partir de ces résultats, des préconisations pour la mise en œuvre industrielle sont proposées. Elles concernent le choix de l'atmosphère de frittage et la teneur en chrome nécessaire à l'optimisation de la microstructure.

Social and environmental justice for a growing and developing global population requires significant increases in energy use. A possible means of contributing to this energy increase is to incinerate plutonium from spent fuel of pressurised light water reactors (Pu(PWR)) in high-temperature reactors such as the Pebble Bed Modular Reactor Demonstration Power Plant 400 MWth (PBMR-DPP-400). Previous studies showed that at low temperatures a 3 g Pu(PWR) loading per fuel sphere or less had a positive uniform temperature reactivity coefficient (UTC) in a PBMR DPP-400. The licensing of this fuel design is consequently unlikely. In the present study it was shown by diffusion simulations of the neutronics, using VSOP-99/05, that there is a fuel design containing thorium and plutonium that achieves a negative maximum UTC. Further, a fuel design containing 12 g Pu(PWR) loading per fuel sphere achieved a negative maximum UTC as well as the other PBMR (Ltd.) safety limits of maximum power per fuel sphere, fast fluence and maximum temperatures. It is proposed that the low average thermal neutron flux, caused by reduced moderation and increased absorption of thermal neutrons due to the higher plutonium loading, is responsible for these effects. However, to fully understand the mechanisms involved a detailed quantitative analysis of the roll of each factor is required. A 12 g Pu(PWR) loading per fuel sphere analysis shows a burn-up of 180.7 GWd/tHM which is approximately double the proposed PBMR (Ltd.) low enriched uranium fuel burn-up. The spent fuel has only a decrease of 24.5 % in the Pu content which is sub-optimal with respect to proliferation and waste disposal objectives. Incinerating Pu(PWR) in the PBMR-DPP 400 MWth is potentially licensable and economically feasible and should be considered for application by industry. / MIng (Nuclear Engineering), North-West University, Potchefstroom Campus, 2012

Social and environmental justice for a growing and developing global population requires significant increases in energy use. A possible means of contributing to this energy increase is to incinerate plutonium from spent fuel of pressurised light water reactors (Pu(PWR)) in high-temperature reactors such as the Pebble Bed Modular Reactor Demonstration Power Plant 400 MWth (PBMR-DPP-400). Previous studies showed that at low temperatures a 3 g Pu(PWR) loading per fuel sphere or less had a positive uniform temperature reactivity coefficient (UTC) in a PBMR DPP-400. The licensing of this fuel design is consequently unlikely. In the present study it was shown by diffusion simulations of the neutronics, using VSOP-99/05, that there is a fuel design containing thorium and plutonium that achieves a negative maximum UTC. Further, a fuel design containing 12 g Pu(PWR) loading per fuel sphere achieved a negative maximum UTC as well as the other PBMR (Ltd.) safety limits of maximum power per fuel sphere, fast fluence and maximum temperatures. It is proposed that the low average thermal neutron flux, caused by reduced moderation and increased absorption of thermal neutrons due to the higher plutonium loading, is responsible for these effects. However, to fully understand the mechanisms involved a detailed quantitative analysis of the roll of each factor is required. A 12 g Pu(PWR) loading per fuel sphere analysis shows a burn-up of 180.7 GWd/tHM which is approximately double the proposed PBMR (Ltd.) low enriched uranium fuel burn-up. The spent fuel has only a decrease of 24.5 % in the Pu content which is sub-optimal with respect to proliferation and waste disposal objectives. Incinerating Pu(PWR) in the PBMR-DPP 400 MWth is potentially licensable and economically feasible and should be considered for application by industry. / MIng (Nuclear Engineering), North-West University, Potchefstroom Campus, 2012

The predominant nuclear fuel used globally at present is uranium which is a finite resource. Thorium has been identified as an alternative nuclear fuel source that can be utilized in almost all existing uranium–based reactors and can significantly help in conserving limited uranium reserves. Furthermore, the elimination of proliferation risks associated with thorium–based fuel cycles is a key reason for re–evaluating the possible utilization of thorium in high temperature reactors. In addition to the many advantages that thorium–based fuel has over uranium–based fuel, there are vast thorium resources in the earth's crust that up until the present have not been exploited optimally. This study focuses on determining the amount of uranium ore that can be saved using thorium as a nuclear fuel in HTR's. Four identical 200 MWth high temperature reactors are considered which make use of different fuel cycles. These fuel cycles range from the conventional uranium fuel cycle to a thorium–based fuel cycle in which no U–238 is present and have been simulated using the VSOP–A system of computer codes. This study also considers the effect that protactinium, an isotope that occurs in thorium–based fuel cycles, will have on the decay heat production in the case of a depressurized loss of coolant (DLOFC) accident. / Thesis (M.Ing. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2011.

The predominant nuclear fuel used globally at present is uranium which is a finite resource. Thorium has been identified as an alternative nuclear fuel source that can be utilized in almost all existing uranium–based reactors and can significantly help in conserving limited uranium reserves. Furthermore, the elimination of proliferation risks associated with thorium–based fuel cycles is a key reason for re–evaluating the possible utilization of thorium in high temperature reactors. In addition to the many advantages that thorium–based fuel has over uranium–based fuel, there are vast thorium resources in the earth's crust that up until the present have not been exploited optimally. This study focuses on determining the amount of uranium ore that can be saved using thorium as a nuclear fuel in HTR's. Four identical 200 MWth high temperature reactors are considered which make use of different fuel cycles. These fuel cycles range from the conventional uranium fuel cycle to a thorium–based fuel cycle in which no U–238 is present and have been simulated using the VSOP–A system of computer codes. This study also considers the effect that protactinium, an isotope that occurs in thorium–based fuel cycles, will have on the decay heat production in the case of a depressurized loss of coolant (DLOFC) accident. / Thesis (M.Ing. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2011.

Title: Variations of actinide magnetism in uranium-base hydrides and other selected systems. Author: Anna Maria Adamska Department / Institute: Department of Condensed Matter Physics Supervisor of the doctoral thesis: Doc. RNDr. Ladislav Havela, CSc., the Department of Condensed Matter Physics Faculty of Mathematics and Physics Charles University, Prague, Czech Republic and Assoc. Prof. Dr. hab. Nhu-Tarnawska Hoa Kim Ngan, Institute of Physics, Pedagogical University, Kraków, Poland. Abstract: Actinide magnetism was studied in three different types of systems. Variations of magnetic properties of UTGe hydrides as a function of hydrogen concentration prove that doping of U intermetallics by interstitial hydrogen leads to stronger magnetic properties, primarily caused by an increase of the inter-uranium separation. Sputter-deposited UFe2+x films, which are derived from the UFe2 Laves phase but have an amorphous structure, exhibit an increase of the Curie temperarture (to more than 400 K) with the Fe excess, which could not be achieved in the bulk. This is understood as a result of the prominent role of the 3d magnetism of Fe. Notoriously weakly magnetic plutonium was studied in the form of the alloy in the ζ-phase, which exists between 35 and 70 % U in Pu. Its susceptibility increases in a comparison to pure...

Sulfate-reducing bacteria (SRB) frequently occur in the deep granitic rock aquifers at the Äspö Hard Rock Laboratory (Äspö HRL), Sweden. The new SRB strain Desulfovibrio äspöensis could be iso-lated. The objective of this project was to explore the basic interaction mechanisms of uranium, curium, neptunium and plutonium with cells of D. äspöensis DSM 10631T. The cells of D. äspöensis were successfully cultivated under anaerobic conditions as well in an optimized bicarbonate-buffered mineral medium as on solid medium at 22 °C. To study the interaction of D. äspöensis with the actinides, the cells were grown to the mid-exponential phase (four days). The collected biomass was usually 1.0±0.2 gdry weight/L. The purity of the used bacterial cultures was verified using microscopic techniques and by applying the Amplified Ribosomal DNA Restriction Enzyme Analysis (ARDREA). The interaction experiments with the actinides showed that the cells are able to remove all four actinides from the surrounding solution. The amount of removed actinide and the interaction mechanism varied among the different actinides. The main U(VI) removal occurred after the first 24 h. The contact time, pH and [U(VI)]initial influence the U removal efficiency. The presence of uranium caused a damaging of the cell membranes. TEM revealed an accumulation of U inside the bacterial cell. D. äspöensis are able to form U(IV). A complex interaction mechanism takes place consisting of biosorption, bioreduction and bioaccumulation. Neptunium interacts in a similar way. The experimental findings are indicating a stronger interaction with uranium compared to neptunium. The results obtained with 242Pu indicate the ability of the cells of D. äspöensis to accumulate and to reduce Pu(VI) from a solution containing Pu(VI) and Pu(IV)-polymers. In the case of curium at a much lower metal concentration of 3x10-7 M, a pure biosorption of Cm(III) on the cell envelope forming an inner-sphere surface complex most likely with organic phosphate groups was detected. To summarize, the strength of the interaction of D. äspöensis with the selected actinides at pH 5 and actinide concentrations ≥10 mg/L ([Cm] 0.07 mg/L) follows the pattern: Cm > U > Pu >> Np.

Objective of this project was the study of basic interaction and migration processes of actinides in the environment in presence of humic acids (HA). To obtain more basic knowledge on these interaction processes synthetic HA with specific functional properties as well as 14C-labeled HA were synthesized and applied in comparison to the natural HA Aldrich. One focus of the work was on the synthesis of HA with distinct redox functionalities. The obtained synthetic products that are characterized by significantly higher Fe(III) redox capacities than Aldrich HA were applied to study the redox properties of HA and the redox stability of U(VI) humate complexes. It was confirmed that phenolic OH groups play an important role for the redox properties of HA. However, the results indicate that there are also other processes than the single oxidation of phenolic OH groups and/or other functional groups contributing to the redox behavior of HA. A first direct-spectroscopic proof for the reduction of U(VI) by synthetic HA with distinct redox functionality was obtained. The complexation behavior of synthetic and natural HA with actinides (Th, Np, Pu) was studied. Structural parameters of Pu(III), Th(IV), Np(IV) and Np(V) humates were determined by X-ray absorption spectroscopy (XAS). The results show that carboxylate groups dominate the interaction between HA and actinide ions. These are predominant monodentately bound. The influence of phenolic OH groups on the Np(V) complexation by HA was studied with modified HA (blocked phenolic OH groups). The blocking of phenolic OH groups induces a decrease of the number of maximal available complexing sites of HA, whereas complex stability constant and Np(V) near-neighbor surrounding are not affected. The effects of HA on the sorption and migration behavior of actinides was studied in batch and column experiments. Th(IV) sorption onto quartz and Np(V) sorption onto granite and its mineral constituents are affected by the pH value and the presence of HA. HA exhibits a significant influence on the transport of U(IV) and U(VI) in a laboratory quartz sand system. In order to provide the basis for a more reliable modeling of the actinide transport, the metal ion complexation with HA has to be integrated into existing geochemical speciation codes. Within this project the metal ion charge neutralization model was embedded into the geochemical modeling code EQ3/6. In addition to that, a digital data base was developed which covers HA complexation data basing on the charge neutralization model.

En raison de leurs propriétés chimiques et physiques, les oxydes mixtes d'uranium et de plutonium sont considérés comme combustibles pour les réacteurs nucléaires de quatrième génération. Dans ce cadre, des études expérimentales complémentaires sont nécessaire, notamment pour mieux comprendre les phénomènes mis en jeu lors de la fabrication ou sous irradiation. L'objet de ce travail est d'étudier le diagramme de phase U-Pu-O dans une large gamme de composition et de températures afin d'améliorer notre connaissance de ce système. La plupart des expériences ont été réalisées par diffraction des rayons X en fonction de la température. La contrôle in situ de la pression partielle en oxygène a permis de faire varier la stœchiométrie en oxygène dans le matériau. L'approche expérimentale a été couplée avec la modélisation thermodynamique par la méthode CALPHAD afin de mieux dimensionner les expériences et interpréter les résultats. Cette méthodologie a permis d'améliorer notre connaissance des équilibres de phase dans le système U-Pu-O. / Due to their physical and chemical properties, mixed uranium-plutonium oxides are considered for fuel in 4th generation nuclear reactors. In this frame, complementary experimental studies are necessary to develop a better understanding of the phenomena that take place during fabrication and operation in the reactor. The focus of this work was to study the U Pu–O phase diagram in a wide range of compositions and temperatures to ameliorate our knowledge of the phase equilibria in this system. Most of experiments were done using in situ X-ray diffraction at elevated temperatures. The control of the oxygen partial pressure during the treatments made it possible to change the oxygen stoichiometry of the sample, which gave us an opportunity to study rapidly different compositions and the processes involved. The experimental approach was coupled with thermodynamic modeling using the CALPHAD method, to precisely plan the experiments and interpret the obtained results. This approach enabled us to enhance the knowledge of phase equilibria in the U–Pu–O system.

Le code MURE est basé sur le couplage d’un code Monte Carlo statique et le calcul de l’évolution pendant l’irradiation et les différentes périodes du cycle (refroidissement, fabrication). Le code MURE est ici utilisé pour analyser deux différentes questions : le mono-recyclage de l’Am dans les réacteurs français de type REP et la conversion du coeur du MNSR (Miniature Neutron Source Reactor) au Ghana d’un combustible à uranium hautement enrichi (HEU) vers un combustible faiblement enrichi (LEU), dans le cadre de la lutte contre la prolifération. Dans les deux cas, une comparaison détaillée est menée sur les taux d’irradiation et les radiotoxicités induites (combustibles usés, déchets).Le combustible UOX envisagé est enrichi de telle sorte qu’il atteigne un taux d’irradiation de 46 GWj/t et 68 GWj/t. Le combustible UOX usé est retraité, et le retraitement standard consiste à séparer le plutonium afin de fabriquer un combustible MOX sur base d’uranium appauvri. La concentration du Pu dans le MOX est déterminée pour atteindre un taux d’irradiation du MOX de 46 et 68 GWj/t. L’impact du temps de refroidissement de l’UOX usé est étudié (5 à 30 ans), afin de quantifier l’impact de la disparition du 241PU (fissile) par décroissance radioactive (T=14,3 ans). Un refroidissement de 30 ans demande à augmenter la teneur en Pu dans le MOX. L’241Am, avec une durée de vie de 432 ans, jour un rôle important dans le dimensionnement du site de stockage des déchets vitrifiés et dans leur radiotoxicité à long terme. Il est le candidat principal à la transmutation, et nous envisageons donc son recyclage dans le MOX, avec le plutonium. Cette stratégie permet de minimiser la puissance résiduelle et la radiotoxicité des verres, en laissant l’Am disponible dans les MOX usés pour une transmutation éventuelle future dans les réacteurs rapides. Nous avons étudié l’impact neutronique d’un tel recyclage. Le temps de refroidissement de l’UOX est encore plus sensible ici car l’241Am recyclé est un fort poison neutronique qui dégrade les performances du combustible (taux d’irradiation, coefficients de vide et de température). Néanmoins, à l’exception de quelques configurations, le recyclage de l’Am ne dégrade pas les coefficients de sûreté de base. Le réacteur MNSR du Ghana fonctionne aujourd’hui avec de l’uranium enrichi à 90,2% (HEU), et nous étudions ici la possibilité de le faire fonctionner avec de l’uranium enrichi à 12,5%, en passant d’un combustible sur base d’aluminium à un oxyde. Les simulations ont été menées avec le code MURE, et montrent que le coeur LEU peut-être irradié plus longtemps, mais demande d’intervenir plus tôt sur le pilotage en jouant sur la quantité de béryllium en coeur. Les flux de neutrons dans les canaux d’irradiation sont similaires pour les coeurs HEU et LEU, de même pour les coefficients de vide. Le combustible LEU usé présente cependant une radiotoxicité et une chaleur résiduelle plus élevée, du fait de la production plus importante de transuraniens pendant l’irradiation. / The MURE code is based on the coupling of a Monte Carlo static code and the calculation of the evolution of the fuel during irradiation and cooling periods. The MURE code has been used to analyse two different questions, concerning the mono-recycling of Am in present French Pressurized Water Reactor, and the conversion of high enriched uranium (HEU) used in the Miniature Neutron Source Reactor in Ghana into low enriched uranium (LEU) due to proliferation resistance issues. In both cases, a detailed comparison is made on burnup and the induced radiotoxicity of waste or spent fuel. The UOX fuel assembly, as in the open cycle system, was designed to reach a burn-up of 46GWd/T and 68GWd/T. The spent UOX was reprocessed to fabricate MOX assemblies, by the extraction of Plutonium and addition of depleted Uranium to reach burn-ups of 46GWd/T and 68GWd/T, taking into account various cooling times of the spent UOX assembly in the repository. The effect of cooling time on burnup and radiotoxicity was then ascertained. Spent UOX fuel, after 30 years of cooling in the repository required higher concentration of Pu to be reprocessed into a MOX fuel due to the decay of Pu-241. Americium, with a mean half-life of 432 years, has high radiotoxic level, high mid-term residual heat and a precursor for other long lived isotope. An innovative strategy consists of reprocessing not only the plutonium from the UOX spent fuel but also the americium isotopes which dominate the radiotoxicity of present waste. The mono-recycling of Am is not a definitive solution because the once-through MOX cycle transmutation of Am in a PWR is not enough to destroy all the Am. The main objective is to propose a “waiting strategy” for both Am and Pu in the spent fuel so that they can be made available for further transmutation strategies. The MOXAm (MOX and Americium isotopes) fuel was fabricated to see the effect of americium in MOX fuel on the burn-up, neutronic behavior and on radiotoxicity. The MOXAm fuel showed relatively good indicators both on burnup and on radiotoxicity. A 68GWd/T MOX assembly produced from a reprocessed spent 46GWd/T UOX assembly showed a decrease in radiotoxicity as compared to the open cycle. All fuel types understudy in the PWR cycle showed good safety inherent feature with the exception of the some MOXAm assemblies which have a positive void coefficient in specific configurations, which could not be consistent with safety features. The core lifetimes of the current operating 90.2% HEU UAl fuel and the proposed 12.5% LEU UOX fuel of the MNSR were investigated using MURE code. Even though LEU core has a longer core life due to its higher core loading and low rate of uranium consumption, the LEU core will have it first beryllium top up to compensate for reactivity at earlier time than the HEU core. The HEU and LEU cores of the MNSR exhibited similar neutron fluxes in irradiation channels, negative feedback of temperature and void coefficients, but the LEU is more radiotoxic after fission product decay due to higher actinides presence at the end of its core lifetime.