The goal of this research was to identify and fabricate zeolitic membranes that can separate radioisotope krypton-85 (half-life 10.72 years) and xenon gas released during spent nuclear fuel reprocessing. In spent nuclear fuel reprocessing, fissionable plutonium and uranium are recovered from spent nuclear fuel and recycled. During the process, krypton-85 and xenon are released from the spent nuclear fuel as process off-gas. The off-gas also contains NO, NO2, 129I, 85Kr, 14CO2, tritium (as 3H2O), and air and is usually vented to the atmosphere as waste without removing many of the radioactive components, such as 85Kr. Currently, the US does not reprocess spent nuclear fuel. However, as a member of the International Framework for Nuclear Energy Cooperation (IFNEC, formerly the Global Nuclear Energy Partnership), the United States has partnered with the international nuclear community to develop a “closed” nuclear fuel cycle that efficiently recycles all used nuclear fuel and safely disposes all radioactive waste byproducts. This research supports this initiative through the development of zeolitic membranes that can separate 85Kr from nuclear reprocessing off-gas for capture and long-term storage as nuclear waste. The implementation of an 85Kr/Xe separation step in the nuclear fuel cycle yields two main advantages. The primary advantage is reducing the volume of 85Kr contaminated gas that must be stored as radioactive waste. A secondary advantage is possible revenue generated from the sale of purified Xe. This research proposed to use a zeolitic membrane-based separation because of their molecular sieving properties, resistance to radiation degradation, and lower energy requirements compared to distillation-based separations. Currently, the only commercial process used to separate Kr and Xe is cryogenic distillation. However, cryogenic distillation is very energy intensive because the boiling points of Kr and Xe are -153 °C and -108 °C, respectively. The 85Kr/Xe separation step was envisioned to run as a continuous cross-flow filtration process (at room temperature using a transmembrane pressure of about 1 bar) with a zeolite membrane separating krypton-85 into the filtrate stream and concentrating xenon into the retentate stream. To measure process feasibility, zeolite membranes were synthesized on porous α-alumina support discs and permeation tested in dead-end filtration mode to measure single-gas permeance and selectivity of CO2, CH4, N2, H2, He, Ar, Xe, Kr, and SF6. Since the kinetic diameter of krypton is 3.6 Å and xenon is 3.96 Å, zeolites SAPO-34 (pore size 3.8 Å) and DDR (pore size 3.6 Å) were studied because their pore sizes are between or equal to the kinetic diameters of krypton and xenon; therefore, Kr and Xe could be separated by size-exclusion. Also, zeolite MFI (average pore size 5.5 Å) permeance and selectivity were evaluated to produce a baseline for comparison, and amorphous carbon membranes (pore size < 5 Å) were evaluated for Kr/Xe separation as well. After permeation testing, MFI, DDR, and amorphous carbon membranes did not separate Kr and Xe with high selectivity and high Kr permeance. However, SAPO-34 zeolite membranes were able to separate Kr and Xe with an average Kr/Xe ideal selectivity of 11.8 and an average Kr permeance of 19.4 GPU at ambient temperature and a 1 atm feed pressure. Also, an analysis of the SAPO-34 membrane defect permeance determined that the average Kr/Xe selectivity decreased by 53% at room temperature due to unselective defect permeance by Knudsen diffusion. However, sealing the membrane defects with polydimethylsiloxane increased Kr/Xe selectivity by 32.8% to 16.2 and retained a high Kr membrane permeance of 10.2 GPU at ambient temperature. Overall, this research has shown that high quality SAPO-34 membranes can be consistently fabricated to achieve a Kr/Xe ideal selectivity >10 and Kr permeance >10 GPU at ambient temperature and 1 atm feed pressure. Furthermore, a scale-up analysis based on the experimental results determined that a cross-flow SAPO-34 membrane with a Kr/Xe selectivity of 11.8 and an area of 4.2 m2 would recover 99.5% of the Kr from a 1 L/min feed stream containing 0.09% Kr and 0.91% Xe at ambient temperature and 1 atm feed pressure. Also, the membrane would produce a retentate stream containing 99.9% Xe. Based on the SAPO-34 membrane analysis results, further research is warranted to develop SAPO-34 membranes for separating 85Kr and Xe.

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One of the proposed solutions to the issue of nuclear waste volume is to transition from once through nuclear fuel cycle to advanced fuel cycles with used fuel recycling option. In any advanced fuel cycles with recycling options, the type and amount of separation technology deployed play a crucial role in the overall performance of the fuel cycle. In this work, a scenario study involving two advanced fuel cycles in addition to the once through fuel cycle were evaluated using VISION nuclear fuel cycle simulation code. The advanced fuel cycles were setup to transition completely to full recycling without any light water reactor by assuming all LWR currently in operation will have 20 years of operating life extension and no new LWR will be constructed thereafter. Several different separation capacities (1kT/yr, 2kT/yr and 4 kT/yr) were deployed and the overall impact of these capacities was analyzed in terms of resources utilization, used fuel and waste material generated and the amount of storage space required. Economic parameter (LCOE, LFCC, etc) analysis was also performed using VISION.ECON. Results presented in this work suggest that the need for LWR-UNF storage can be minimized if sufficient separation capacity is deployed early in the fuel cycle. It can also be concluded that a FuRe system without LEU will not be feasible, thus SFRs must be designed for optional use of LEU fuel. Otherwise LWRs must continue to be part of the mix to keep the near term cost of generating electricity competitive. It was observed that the higher amount of separation capacity deployed in the advanced fuel cycles led to higher LFCC and LCOE, but also translates into less environmental impact on both front and back end of the fuel cycle.

Les isotopes de l’Américium sont les principaux contributeurs à la radioactivité des déchets nucléaires. Parmi les scénarios pour diminuer la toxicité des déchets, la transmutation dans les réacteurs à neutrons rapides utilisant des pastilles d’oxyde mixte (U,Am)O2 est une voie prometteuse. Dans ce cadre, la connaissance des propriétés thermodynamiques du système U-Am-O est essentielle pour prédire le comportement des pastilles (U,Am)O2 en conditions nominale et accidentelle. Cette thèse est dédiée à l’étude expérimentale d’oxydes mixtes (U,Am)O2 dans une large gamme de composition (7,5 % at. ≤ Am/(Am+U) ≤ 70 % at.). L’objectif est d’acquérir des données pour développer un modèle thermodynamique avec la méthode semi-empirique CALPHAD. Les résultats peuvent être classés en trois catégories : données structurales, données de diagramme de phase et données thermodynamiques. Pour la modélisation thermodynamique d’un système ternaire, l’optimisation des sous-systèmes binaires est nécessaire. Comme des questions restaient en suspens sur le système Am-O, le diagramme de phase Am-O a tout d’abord été étudié par diffraction des rayons X à haute température. L’existence d’un domaine de composition de la phase bcc AmO1.61 a été mis en évidence et la lacune de miscibilité dans la phase fluorite, proposée dans la littérature, n’a pas été confirmée. Grâce à ces nouveaux résultats, le modèle CALPHAD de Gotcu et al a été modifié. Dans une deuxième étape, des analyses structurales des dioxydes (U,Am)O2±x ont été effectuées par DRX, XAS et spectroscopie RAMAN. La DRX a permis de confirmer que tous les échantillons sont constitués d’une seule phase de structure fluorite. Le rapport O/M (avec M=U+Am) mesuré à température ambiante est inférieur à 2 ; la stabilité de l’Américium trivalent Am3+ a été mise en évidence. Celle-ci induit l’oxydation partielle de l’U4+ en U5+. Cette distribution de charge s’accompagne par la formation de défauts de l’oxygène complexes dans la structure fluorite. Lors de l’étude par DRX HT des oxydes mixtes sous air, il a été montré que la présence d’Am3+ stabilise la phase fluorite par rapport aux oxydes plus riches en oxygène (U4O9, U3O8). De nouvelles données de diagramme de phase ont été obtenues : des conodes dans les domaines biphasés M4O9-M3O8 and MO2+x-M3O8 et la solubilité de l’Américium dans les oxydes M4O9 et M3O8. L’étude du diagramme de phase U-Am-O a été poursuivie par la détermination des températures de solidus/liquidus des oxydes mixtes par une technique de chauffage laser, sous argon et sous air, et par la caractérisation des échantillons après fusion par SEM et XAS. La température de fusion des oxydes mixtes diminue avec une teneur croissante d’Américium (Am/(Am+U)) et d’oxygène (O/(Am+U)). Finalement, les propriétés thermodynamiques des oxydes (U,Am)O2±x ont été mesurées : les incréments enthalpiques par calorimétrie de chute et les pressions partielles des espèces gazeuses par Spectrométrie de Masse couplée à une cellule de Knudsen (KEMS). Une contribution d’excès de la capacité calorifique a été observée à haute température, attribuée à la réduction des oxydes (avec formation de lacunes d’oxygène). Les résultats de KEMS ont permis de déterminer une composition congruente de vaporisation à 2300 K, pour un rapport Am/(Am+U) de 0,6 et un rapport O/(U+Am) inférieur à 1,9. Finalement, la modélisation thermodynamique du système U-Am-O par la méthode CALPHAD a été abordée par la description de la phase fluorite. Un bon accord est obtenu entre le modèle et les données de potentiel d’oxygène pour l’oxyde (U0.5Am0.5)O2±x et de distribution des cations. De plus, le modèle permet de reproduire de façon satisfaisante les données de KEMS. En perspective de ce travail, la modélisation thermodynamique du ternaire sera étendue à la description des équilibres de phase mettant en jeu les oxydes M4O9, M3O8 et la phase liquide. / Americium isotopes are the main contributors to the long-term radiotoxicity of the nuclear wastes, after the plutonium extraction. Among the reprocessing scenarios, the transmutation in fast neutron reactors using uranium-americium mixed oxide (U,Am)O2±x pellets seems promising. In this frame, the knowledge of the thermodynamics of the U-Am-O ternary system is of essential for the prediction of the behavior of (U,Am)O2 pellets and their possible interaction with the cladding, under normal and accidental conditions. This thesis is dedicated to the experimental investigation of U-Am mixed oxides on a wide range of Am contents (7.5 at.% ≤ Am/(Am+U) ≤ 70 at.%), with the aim to collect data for developing a thermodynamic model based on the semi-empirical CALPHAD method. The obtained results can be classified in three categories: structural, phase diagram and thermodynamic data. For the thermodynamic modeling of the ternary system, the assessment of the binary sub-systems is first required. As open questions still existed on the Am-O system, a first part of the work was dedicated to the study of the Am-O phase diagram by high-temperature (HT) XRD. The existence of a composition range of the bcc AmO1.61 phase was highlighted and the miscibility gap in the fluorite phase, proposed in the literature, was not found. Thanks to the new experimental data, the existing CALPHAD model of Gotcu et al. was modified. In a second step, structural investigations were performed on synthesized (U,Am)O2±x dioxides by coupling XRD, XAS and Raman spectroscopy. For all the compositions, the XRD confirmed the formation of a single fluorite structure. The O/M ratio (with M=U+Am) at room temperature was determined to be lower than 2; the stability of trivalent americium Am3+ in the dioxide solid solution was highlighted, which induces a partial oxidation of uranium from U4+ to U5+. This charge distribution, peculiar for a dioxide, is accompanied by the formation of complex oxygen defects in the fluorite structure. By a HT-XRD investigation of the mixed oxides under air combined with XAS characterization of the oxidized samples, it was shown that the presence of Am3+ leads to a stabilization of the dioxide fluorite phase toward the formation of oxides richer in oxygen, in comparison to the U-O system. New phase diagram data were obtained in the oxygen rich region at 1470 K: tie-lines in the M4O9-M3O8 and MO2+x-M3O8 domains were determined and the solubility of americium in the M4O9 and M3O8 oxides was estimated. The investigation of the U-Am-O phase diagram continued at higher temperature with the study of the solidus/liquidus transitions using a laser-heating technique, under argon and air, and post-melting characterizations conducted by SEM and XAS. The melting temperature of Am-U dioxides decreases with the increase of both the Am/(Am+U) and O/M ratios. Finally, thermodynamic properties of the U1-yAmyO2±x oxides were measured: enthalpy increments using drop calorimetry, partial vapor pressures by Knudsen cell effusion mass spectrometry (KEMS). An excess contribution to the heat capacity at high temperature was observed and this was attributed to the reduction of the dioxides at high temperature (formation of oxygen vacancies). The KEMS results lead to determine the congruent vaporization composition at 2300 K, for a Am/(Am+U) ratio of 0.6 and an O/M ratio lower than 1.9. Finally, the CALPHAD thermodynamic assessment of the U-Am-O system was started, by focusing the attention on the modelling of the fluorite phase. A good agreement between the model and the oxygen potential data for (U0.5Am0.5O2±x) and the cation distribution was achieved. Furthermore, the model is able to satisfactorily reproduce the KEMS data and hence the equilibrium between the dioxide and gas phase. For the perspectives of this work, the optimization of the thermodynamic model should be extended to describe the phase equilibria involving the M4O9, M3O8 oxides and the liquid phase.

Le procédé de traitement du combustible du réacteur à sels fondus (réacteur de génération IV) est un procédé multi-étape dans lequell’extraction des actinides et des lanthanides utilise la technique d’extraction réductrice. Le développement d’un modèle analytique a montré que la mise en contact du sel combustible LiF-ThF4 avec une phase métallique constituée d'un mélange Bi-Li permet l’extraction sélective et quantitative des actinides dans un premier temps, puis l’extraction quantitative des lanthanides dans un second temps. La maitrise de ce procédé nécessite la connaissance des caractéristiques des phases salines impliquées dans le procédé. Les études des propriétés physico-chimiques des sels fluorures fondus ont permis de développer une technique de mesure de la fluoroacidité dans ces milieux via une mesure potentiométrique. Cette technique a permis d’établir un classement de différents mélanges de fluorures fondus en fonction de leur acidité relative. Par ailleurs, une méthode de détermination de la solvatation de solutés dans ces milieux a également été développée par électrochimie afin d’approfondir la connaissance du sel combustible (en particulier solvatation de ThF4 par les ions F-).L'extraction réductrice met également en jeu une phase métallique liquide. Une technique de préparation de cette phase a été développée par électro-réduction de lithium sur une électrode liquide de bismuth en milieu LiCl-LiF. Cette technique permet un bon contrôle de la fraction molaire de lithium introduite dans le bismuth, paramètre essentiel à l’efficacité de l’extraction.Enfin, afin d'optimiser le procédé général de traitement multi-étapes, des méthodes électrochimiques ont été proposées afin de régénérer les différentes phases liquides (salines et métalliques) mise en jeu lors de l’extraction. / The fuel reprocessing of the molten salt reactor (Gen IV concept) is a multi-steps process in which actinides and lanthanides extraction is performed by a reductive extraction technique. The development of an analytic model has showed that the contact between the liquid fuel LiF-ThF4 and a metallic phase constituted of Bi-Li provide firstly a selective and quantitative extraction of actinides and secondly a quantitative extraction of lanthanides. The control of this process implies the knowledge of saline phase properties. Studies of the physico-chemical properties of fluoride salts lead to develop a technique based on potentiometric measurements to evaluate the fluoroacidity of the salts. An acidity scale was established in order to classify the different fluoride salts considered.Another electrochemical method was also developed in order to determine the solvation properties of solutes in fluoride F- environment (and particularly ThF4 by F-)In reductive extraction technique, a metallic phase is also involved. A method to prepare this phase was developed by electro-reduction of lithium on a bismuth liquid cathode in LiCl-LiF melt. This technique allows to accurately control the molar fraction of lithium introduced into the liquid bismuth, which is a main parameter to obtain an efficient extraction.