En France, le cycle actuel de traitement du combustible usé génère un rejet de carbone 14 (déchet à vie longue, période  5730 ans) dans la géosphère. Ce document apporte des premiers éléments scientifiques et techniques pour évaluer la faisabilité d’un conditionnement du carbone, en tant qu’alternative aux rejets. Un matériau de type céramique a été choisi en raison d’un volume spécifique de 2 à 4 L.kg-1 de carbone, contre 12 à 17 L.kg-1 de carbone pour la cimentation actuellement mise en œuvre industriellement à l’étranger, afin de limiter les volumes produits en prévision d’un stockage géologique potentiel. La phase barytocalcite BaCa(CO3)2 a été étudiée en termes de (i) mode de synthèse compatible avec la forme existante du carbone 14 dans le cycle (ii) solidification par traitement thermique et (iii) résistance à la lixiviation. La barytocalcite a pu être obtenue par précipitation et densifiée à plus de 95% avec un frittage SPS de 30 min à 450°C sous 70 MPa typiquement, et ce, sans perte de carbone. Toutefois, le matériau présente une durabilité chimique limitée en eau pure à 30°C mais le carbone relâché précipite sous la forme de BaCO3 et CaCO3. Malgré cela, la potentialité de la barytocalcite en tant que matériau de conditionnement du carbone 14 reste réelle en raison des propriétés de rétention du site de stockage potentiel. Dans une seconde partie, ce document présente des résultats de caractérisation par diffusion et diffraction de rayons X appliqués à des matériaux liés à la R&D sur le traitement de déchets nucléaires. / In France, the current recycling process of the spent nuclear fuel produces a carbon-14 release in the geosphere (long life radionuclide, half-life  5730 y). This document provides scientific and technical data to evaluate the feasibility of carbon conditioning as an alternative to its release. A ceramic type material was chosen because of a specific volume of 2-4 L/kg of carbon, against 12-17 L/kg of carbon for the cementation process currently implemented industrially abroad, to limit the volumes in the case of a potential geological storage. The barytocalcite phase BaCa(CO3)2 has been studied in terms of (i) synthesis process compatible with the existing form of carbon-14 in the cycle (ii) solidification by heat treatment and (iii) leach resistance. The barytocalcite has been obtained by precipitation and densified to more than 95% with SPS sintering for 30 min at 450°C under 70 MPa, without any loss of carbon. This phase has a limited chemical durability in pure water at 30°C but the released carbon ions precipitate as BaCO3 and CaCO3. However the potentiality of barytocalcite as a carbon 14 conditioning matrix remains interesting because of the retention properties of the potential storage site. In a second part, this document provides the results of characterizations made by X-rays scattering and diffraction on different phases in relation with the nuclear waste conditioning R&D.

Glaciation is considered one of the main natural processes that can have a significant impact on the long term performance of DGRs. The northern part of the American continent has been subjected to a series of strong glaciation and deglaciation events over the past million years. Glacial cycles cause loading and unloading, temperature changes and hydraulic head changes at the ground surface. These changes can be classified as transient boundary conditions. It is widely accepted that the periodic pattern of past glacial cycles during the Late Quaternary period are resultant of the Earth’s orbital geometry changes that is expected to continue in the future. Therefore, from the safety perspective of DGRs, such probable events need to be taken into account. The objective of this thesis is to develop a numerical model to investigate the thermo-hydro-mechanical-chemical (THMC) coupled processes that have resulted from long term past and future climate changes and glaciation cycles on a proposed DGR in sedimentary rocks in southern Ontario. The first application is done on a large geological cross section that includes the entire Michigan basin by using a hydro-mechanical (HM) coupled process. The results are compared with field data of anomalous pore water pressures from deep boreholes in sedimentary rocks of southern Ontario. In this work. The modeling results seem to support the hypothesis that at least the underpressures in the Ordovician formation could be partially attributed to past glaciation. The second application is made on site conditions by using the THMC model. The results for the pore water pressure, tracer profiles, permafrost depth and effective stress profile are compared with the available field data, the results show that the solute transport in the natural limestone and shale barrier formations is controlled by diffusion, which provide evidence that the main mechanism of transport at depth is diffusion-dominant. The third application is made on site conditions to determine the effect of underground changes in DGRs due to DGR construction. The results show that future glaciation loads will induce larger increases in effective stresses on the shaft. Furthermore, it is found that hypothetical nuclide transport in a failed shaft can be controlled by diffusion and advection. The simulation results show that the solute transported in a failed shaft can reach the shallow bedrock groundwater zone. These results might imply that a failed shaft will substantially lose its effectiveness as a barrier. The fourth application is proposed to investigate the geochemical evolution of sedimentary host rock in a near field scale. In this part, a new thermo-hydro-mechanical-geochemical simulator (COMSOL-PHREEQC) is developed. It is anticipated that there will be a geochemical reaction within the host rock that results from interaction with the water enriched with the CO2 generated by nuclear waste.

Les Colis Standards de Déchets Compactés (CSD-C) sont des déchets issus du retraitement des assemblages de combustibles nucléaires. Ils sont en partie constitués des gaines oxydées de Zircaloy-4. Ces pièces métalliques sont cisaillées avant d'être placées dans un étui en acier et compactées sous forme de galettes. Ces galettes contiennent des traces de produits d'activation, de produits de fission et d'actinides présents à la surface du Zircaloy-4 oxydé. Dans l'hypothèse d'un éventuel stockage en couche géologique profond, le relâchement des radioéléments contenus dans les CSD-C s'effectuerait après l'altération des pièces métalliques par corrosion au contact de l'eau de re-saturation du site. En effet, cette eau, sous forme vapeur lors de l'entreposage (due à l'humidité résiduelle), puis liquide pendant le stockage sera irradiée. L'irradiation provoque le phénomène de radiolyse de l'eau susceptible d'accélérer les processus de corrosion du Zircaloy-4 oxydé. Cette thèse a pour objectifs de comprendre les mécanismes d'altération du Zircaloy-4 oxydé au contact de l'air humide et de l'eau liquide soumis à des rayonnements ionisants. Nous avons choisi de porter notre attention sur l'impact de la radiolyse induite par irradiations protons et par irradiations gamma. Pour cela, différentes atmosphères gazeuses et différents milieux aqueux ont été utilisés. Pour l'atmosphère gazeuse, nous avons fait varier la pression partielle de vapeur d'eau présente dans un mélange représentatif de l'air. Pour l'eau, l'effet de trois compositions de solutions aqueuses sur le comportement du Zircaloy-4 oxydé a été étudié (eau déminéralisée, eau basique et eau simulant l'eau de re-saturation). Nous avons également fait varier l'énergie déposée dans la solution. Deux comportements distincts ont été mis en évidence dans les conditions expérimentales étudiées. La radiolyse de l'air humide donne lieu à des réactions chimiques en surface du Zircaloy-4 oxydé conduisant à la formation du composé Sn3(OH)4(NO3)2 et du complexe [Zr4 (OH)8 (H2 O) 16]8+ . La radiolyse de l'eau augmente la vitesse de dissolution du Zircaloy-4. Ce phénomène semble s'effectuer par formation de l'ion HZrO−3 à la surface de l'oxyde puis par sa mise en solution. Les vitesses de dissolution dépendent des conditions d'irradiations et de la composition initiale de la solution aqueuse. Elles sont de l'ordre de quelques nanomètres par an à quelques micromètres par an. / Pas de résumé donné.

Après dissolution du combustible et séparation des différents éléments par le procédé PUREX, la majeure partie des produits de fission et des actinides mineurs est calcinée puis vitrifiée dans des verres de conditionnement des déchets nucléaires. Parmi ces produits de fission, certains précipitent et ne sont pas immobilisés dans la phase vitreuse dédiée aux déchets de haute activité à vie longue. Les éléments platinoïdes Pd-Rh-Ru sont insolubles dans le verre nucléaire. En fonction du potentiel d'oxygène imposé par la fritte de verre, ils précipitent sous la forme de phases oxydes complexes ou de composés intermétalliques principalement alliés aux éléments chalcogènes Te et Se. Au contraire, le molybdène reste oxydé lors des dernières étapes du procédé de vitrification. Très réactif vis-à-vis des oxydes constitutifs de la fonte verrière, il forme majoritairement des molybdates. Au cours de cette thèse, la thermodynamique des systèmes chimiques contenant le molybdène (Mo), les éléments platinoïdes Pd-Rh-Ru et les chalcogènes Se-Te ont été étudiés expérimentalement. En parallèle, la thermodynamique de ces systèmes chimiques est également modélisée par la méthode Calphad. L'objectif de cette modélisation est de prédire les phénomènes de cristallisation du molybdène et des platinoïdes observés au cours des étapes de vitrification en fonction de la composition et de la température. Ces modélisations permettent d'effectuer des calculs d'applications en lien avec le procédé industriel de vitrification. / After the dissolution of the used fuel and the separation of several elements by the PUREX process, the high level nuclear wastes composed of fission products and minor actinides are reprocessed and vitrified in nuclear glasses at AREVA La Hague plant. Some of the fission products precipitate : they are not solubilized in the glass matrix. On the one hand, platinoids Pd-Ru-Rh are not soluble in the nuclear glasses. Depending on the oxygen potential, they form complex solid oxyde phases or intermetallic compounds containing chalcogen elements such as selenium and tellurium. On the other hand, the molybdenum forms only oxide phases during the vitrification process. It reacts strongly with the oxide phases present in the glass melt to form mainly molybdate phases. Some of these phases are only temporary formed but other are more stable and can precipitate in the glass matrix when a large amount of molybdenum is supplied. In this thesis, the thermodynamics of the chemical systems containing molybdenum, the platinoid elements Pd-Rh-Ru and the chalcogen elements Se and Te were experimentally investigated. At the same time, these chemical systems were modeled with the Calphad method so as to be able to predict the crystallization phenomena of molybdenum and the platinoids occurring during the vitrification as a function of the composition and the temperature. These modelings are useful to perform application calculations in relation with the vitrification process.

Glaciation is considered one of the main natural processes that can have a significant impact on the long term performance of DGRs. The northern part of the American continent has been subjected to a series of strong glaciation and deglaciation events over the past million years. Glacial cycles cause loading and unloading, temperature changes and hydraulic head changes at the ground surface. These changes can be classified as transient boundary conditions. It is widely accepted that the periodic pattern of past glacial cycles during the Late Quaternary period are resultant of the Earth’s orbital geometry changes that is expected to continue in the future. Therefore, from the safety perspective of DGRs, such probable events need to be taken into account. The objective of this thesis is to develop a numerical model to investigate the thermo-hydro-mechanical-chemical (THMC) coupled processes that have resulted from long term past and future climate changes and glaciation cycles on a proposed DGR in sedimentary rocks in southern Ontario. The first application is done on a large geological cross section that includes the entire Michigan basin by using a hydro-mechanical (HM) coupled process. The results are compared with field data of anomalous pore water pressures from deep boreholes in sedimentary rocks of southern Ontario. In this work. The modeling results seem to support the hypothesis that at least the underpressures in the Ordovician formation could be partially attributed to past glaciation. The second application is made on site conditions by using the THMC model. The results for the pore water pressure, tracer profiles, permafrost depth and effective stress profile are compared with the available field data, the results show that the solute transport in the natural limestone and shale barrier formations is controlled by diffusion, which provide evidence that the main mechanism of transport at depth is diffusion-dominant. The third application is made on site conditions to determine the effect of underground changes in DGRs due to DGR construction. The results show that future glaciation loads will induce larger increases in effective stresses on the shaft. Furthermore, it is found that hypothetical nuclide transport in a failed shaft can be controlled by diffusion and advection. The simulation results show that the solute transported in a failed shaft can reach the shallow bedrock groundwater zone. These results might imply that a failed shaft will substantially lose its effectiveness as a barrier. The fourth application is proposed to investigate the geochemical evolution of sedimentary host rock in a near field scale. In this part, a new thermo-hydro-mechanical-geochemical simulator (COMSOL-PHREEQC) is developed. It is anticipated that there will be a geochemical reaction within the host rock that results from interaction with the water enriched with the CO2 generated by nuclear waste.

With the widening and increasing use of nuclear energy, it is very important to design and build long-term deep geological repositories (DGRs) to manage radioactive waste. The disposal of nuclear waste in deep rock formations is currently being investigated in several countries (e.g., Canada, China, France, Germany, India, Japan and Switzerland). In Canada, a repository for low and intermediate level radioactive waste is being proposed in Ontario’s sedimentary rock formations. During the post-closure phase of a repository, significant quantities of gas will be generated from several processes, such as corrosion of metal containers or microbial degradation of organic waste. The gas pressure could influence the engineered barrier system and host rock and might disturb the pressure-head gradients and groundwater flows near the repository. An increasing gas pressure could also cause damage to the host rock by inducing the development of micro-/macro-cracks. This will further cause perturbation to the hydrogeological properties of the host rock such as desiccation of the porous media, change in degree of saturation and hydraulic conductivity. In this regard, gas generation and migration may affect the stability or integrity of the integrate barriers and threaten the biosphere through the transmitting gaseous radionuclides as long-term contaminants. Thus, from the safety perspective of DGRs, gas generation and migration should be considered in their design and construction. The understanding and modelling of gas migration within the host rock (natural barrier) and the associated potential impacts on the integrity of the natural barrier are important for the safety assessment of a DGR. Therefore, the key objectives of this Ph.D. study include (i) the development of a simulator for coupled modelling of gas migration in the host rock of a DGR for nuclear waste; and (ii) the numerical investigation of gas migration in the host rock of a DGR for nuclear waste in Ontario by using the developed simulator. Firstly, a new thermo-hydro-mechanical-chemical (THMC) simulator (TOUGHREACT-COMSOL) has been developed to address these objectives. This simulator results from the coupling of the well-established numerical codes, TOUGHREACT and COMSOL. A series of mathematical models, which include an elastoplastic-damage model have been developed and then implemented into the simulator. Then, the predictive ability of the simulator is validated against laboratory and field tests on gas migration in host rocks. The validation results have shown that the developed simulator can predict well the gas migration in host rocks. This agreement between the predicted results and the experimental data indicates that the developed simulator can reasonably predict gas migration in DGR systems. The new simulator is used to predict gas migration and its effects in a potential DGR site in Ontario. Valuable results regarding gas migration in a potential DGR located in Ontario have been obtained. The research conducted in this Ph.D. study will provide a useful tool and information for the understanding and prediction of gas migration and its effect in a DGR, particularly in Ontario.

Afin d’éviter le relâchement d’iode 129 (déchet de moyenne activité à vie longue) dans l’environnement, un nouveau matériau incorporant l’iodate dans une hydroxyapatite phosphocalcique a été étudié. Deux méthodes de préparation de ce matériau ont été développées : élaboration par précipitation suivi d’un frittage SPS et élaboration par voie cimentaire. Une quantité pondérale d’iode (taux d’incorporation maximal de 10%mass.) est incorporé uniquement sous forme iodate dans la structure apatitique préparée à partir des deux méthodes d’élaboration. Un monolithe ayant un taux de densification de 88,6 % a été obtenu après mise en forme de poudres précipitées par frittage SPS. Ce matériau présente une résistance à la lixiviation satisfaisante, caractérisée par une vitesse d’altération initiale en eau pure à 50 °C de 10-2 g.m-2.j-1 (comparable à celle d’un verre R7T7 lixivié dans les mêmes conditions) et par une vitesse d’altération résiduelle à 50 °C de 10-5 g.m-2.j-1 dans l’eau souterraine d’un site potentiel de stockage. Dans l’ensemble, ce matériau est un candidat potentiel pour un conditionnement de l’iode radioactif. / In order to avoid the release of 129I (long-lived intermediate-level waste) in the environment, we describe a novel material incorporating iodate in a calcium phosphate based hydroxyapatite. This material is prepared by two synthetic processes: a wet precipitation route followed by a spark plasma sintering and a cementitious route. A high iodine content (with a maximum incorporation rate of 10 wt.%) is reached for both processes, by incorporation of the iodate in the apatitic structure. A monolith with relative density of 88.6% was obtained after shaping of the precipitated powders by spark plasma sintering. This material reveals satisfactory leaching properties, with an initial leaching rate in pure water at 50 °C of 10-2 g.m-2.j-1, and a residual leaching rate at 50 °C of 10-5 g.m-2.j-1 in underground water of potential geological repositories. All in all, this material is a potential candidate for the conditioning of radioactive iodine.

Bentonite is regarded as an ideal geomaterial for the engineering barrier system of a deep geological repository (DGR) where nuclear wastes are disposed, as it has several desirable properties for sealing the nuclear wastes, including low permeability, low diffusion coefficient, high adsorption capacity and proper swelling ability. Nevertheless, gas migration in saturated bentonite may undermine the sealing ability of the geomaterial. Previous experimental studies showed that the gas migration process is accompanied by complex hydromechanical (HM) behaviors, such as gas breakthrough phenomenon, development of preferential pathways, build-up of water pressure and total stress, nearly saturated state after gas injection test, localized consolidation, water exchange between clay matrix and developed fractures and self-sealing process. These experimentally observed behaviors should be properly modelled for conducting a reliable performance assessment for the geomaterial over the lifespan of DGR. In this thesis, two different coupled HM frameworks, i.e., one based on double porosity (DP) concept, referred to as coupled HM-DP framework, and the other on phase field (PF) method, referred to as coupled HM-PF framework, are proposed to simulate the gas migration process in saturated bentonite. For the coupled HM-DP framework, the saturated bentonite is assumed as a superposition of a MAcro-Continuum (MAC) and a MIcro-Continuum (MIC). Two-phase flow is only allowed in the MAC, whereas the MIC is impermeable to both water and gas. Nevertheless, the water can transfer between the MIC and the MAC under the water pressure gap. The first coupled HM model in this framework is based on a double effective stress concept. Mechanical behaviors of the MAC and the MIC are respectively governed by Bishop-type effective stress and Terzaghi’s effective stress. The model can well simulate the evolutions of both gas pressure and gas outflow rate, the water exchange between clay matrix and developed pathways, the high degree of saturation and the consolidation of clay matrix. To account for the development of preferential pathways, the damaging effect has been introduced in the framework. In this improved model, Bishop-type effective stress for the MAC is replaced by the independent stress state variables, i.e., net normal stress and suction, since using the net normal stress is beneficial to simulating tensile failure under high gas pressure. Numerical results showed that the damage-enhanced model can well describe the effect of the development of preferential pathways on the build-up of water pressure and total stress. In addition, the proposed hysteretic models for intrinsic and relative permeabilities make the coupled HM framework more flexible to reproduce the experimental results. To explicitly simulate the development of preferential pathways, a coupled HM-PF framework is developed by using Coussy’s thermodynamic theory and the microforce balance law. The coupled HM-PF framework is implemented in the standard Finite Element Method (FEM). To avoid the pore pressure oscillation and enhance the computational efficiency, a stabilized mixed finite element, in which linear shape functions are selected for interpolating all primary variables, is adopted to discretize the whole domain. In the developed framework, swelling pressure (initial stress) is accounted for by introducing a modified strain tensor that is the sum of the strain tensor due to deformation and the strain tensor calculated from the initial stress. The numerical results showed that the developed coupled HM-PF framework can capture some important behaviors, such as the discrete pathways, localized gas flow, built-up of water pressure and total stress under constant volume condition and nearly saturated state in clay matrix. A spatially autocorrelated random field is introduced into the framework to describe the heterogeneous distribution of HM properties in bentonite. The heterogeneity is beneficial to simulating the fracture branching and the complex fracture trajectory. Numerical results showed that some factors, such as Gaussian random field, coefficient of variation, boundary condition and injection rate, have significant influences on the fracture trajectory. At the end of the thesis, the obtained numerical results are synthesized and analyzed. Based on the analysis, the pros and cons of the developed numerical models are discussed. Corresponding to the limitations, some recommendations are proposed for future studies.

Ce travail porte sur le comportement sous irradiation d’une vitrocéramique élaborée par traitement thermique d’une version riche en molybdène du verre de confinement de déchets nucléaires R7/T7 et plus particulièrement de la phase cristalline. Des terres rares (Nd3+ et Eu3+) sont utilisées à la fois comme simulants des produits de fission et des actinides mineurs et comme sondes structurales luminescentes. La phase cristalline présente dans ce type de vitrocéramique est un molybdate de calcium de type powellite CaMoO4 ayant incorporé divers éléments dont des terres rares. Les propriétés cristallochimiques de la phase powellite ont été étudiées notamment par spectroscopie Raman et photoluminescence grâce à divers échantillons naturels et céramiques de compositions allant de CaMoO4 à une modèle proche de celle des cristaux de la vitrocéramique : Ca0,76Sr0,1Na0,07Eu0,01La0,02Nd0,02Pr0,02MoO4. La largeur à mi-hauteur de la bande Raman à 880 cm-1 augmente linéairement en fonction du taux d’incorporation sur le site calcium, incorporation qui influe également sur les paramètres de maille. Le volume intrinsèque de la maille augmente ainsi de 2 %. L’étude d’analogues naturels contenant de l’uranium ainsi que de céramiques et vitrocéramiques irradiées aux ions hélium, argon et plomb a permis de montrer que la structure powellite était très résistante aux dégâts causés. Sous irradiation, le signal de luminescence de Eu3+ des différents échantillons tend à s’uniformiser. Cette uniformisation du signal se retrouve également en spectroscopie Raman. Alors qu’elle peut varier entre 6 et 12 cm-1 pour des céramiques saines, la largeur à mi-hauteur de la bande Raman à 880 cm-1 devient identique, de l’ordre de 18 cm-1, à partir de 10 dpa. Le désordre créé par les irradiations prend le pas sur celui créé par l’incorporation d’éléments dans la structure. Cependant, la spectroscopie Raman et la diffraction des rayons X montre que la structure reste cristalline, au moins partiellement. Sous irradiations, le gonflement de la powellite est en moyenne de 5 % mais est très hétérogène. / This work deals with the behaviour under irradiation of a glass-ceramic made after heat treatment of a molybdenum rich R7/T7 type glass. Rare earth elements (Eu3+ and Nd3+) are used as surrogates of minor actinides and fission products as well as structural luminescent probes. We will focus on the behaviour of the crystalline phase which is a powellite type calcium molybdate that incorporated other elements including rare earth elements. In order to determine the crystalline-chemical properties of the powellite structure, Raman spectroscopy and photoluminescence analyses are led on natural powellite samples and synthetic ceramics with compositions from pure CaMoO4 to Ca0.76Sr0.1Na0.07Eu0.01La0.02Nd0.02Pr0.02MoO4, a model composition of the crystalline phase of the glass-ceramic. The analyses of synthetic samples irradiated with He, Ar and Pb ions compared to the behaviour of a natural powellite sample that contains uranium indicate that powellite resist strongly to irradiation and never reach the amorphous state.

La détermination de la concentration des RadioNucléides à Vie Longue (RNVL) dans les déchets de l'industrie nucléaire est essentielle pour la gestion sur le long terme des sites de stockages. Cette étude se focalise sur la détermination de la concentration de trois RNVL : 129I, 41Ca et 10Be dans les résines échangeuses d'ions utilisées pour la purification du fluide primaire des Réacteurs à Eau Pressurisée (REP). Afin d'exploiter les potentialités de la Spectrométrie de Masse par Accélérateur (SMA) pour mesurer ces radionucléides présents en de très faibles concentrations, des procédures analytiques spécifiques ont été développées incluant : 1) la minéralisation des échantillons, 2) l'extraction sélective des analytes, 3) le conditionnement pour la mesure par SMA. Appliquées à des échantillons de résines usées provenant d'une centrale EDF (REP 900 MWe), les procédures développées ont permis l'extraction quantitative et sélective des RNVL d'intérêt vis-à-vis des émetteurs β-γ et des isobares avant leur mesure par SMA sur l'instrument national ASTER (CEREGE, Aix-en-Provence). L'iode 129, le calcium 41 et le béryllium 10 ont été mesurés dans les résines usées à des concentrations de l'ordre de 10 ng/g, 20 pg/g et 4 ng/g de résine sèche, respectivement. Pour ce qui concerne l'iode 129 et le calcium 41, ces concentrations sont en accord avec celles estimées à partir de facteurs de corrélation établis relativement à des émetteurs gamma facilement mesurables (137Cs et 60Co). Dans le cas du béryllium 10, les résultats obtenus différent significativement des valeurs attendues mais sont cohérents avec de précédentes mesures réalisées par ICP-MS. / Determining the concentration of Long-Lived RadioNuclides (LLRN) in nuclear waste is fundamental for the long term management of storage sites. This study focuses on the determination of three LLRN concentrations, i.e. 129I, 41Ca and 10Be, in ion exchange resins used for primary fluid purification in Pressurized Water Reactors (PWR). To benefit from the Accelerator Mass Spectrometry (AMS) technique allowing to measure extremely low levels of nuclide concentrations, analytical procedures including: 1) sample dissolution; 2) selective and quantitative extraction of the analyte; and, 3) analyte conditioning for AMS measurements, were developed. Applied on spent resin samples collected at a 900 MW PWR, the procedures developed for each studied LLRN allowed their quantitative recovery and their selective extraction from β-γ emitters and isobars. The concentration measurements of the LLRN of interest were then performed on the Accelerator Mass Spectrometry national facility ASTER housed by the Centre Européen de Recherche et d'Enseignement des Géosciences de l'Environnement (CEREGE, Aix-en-Provence). 129I, 41Ca and 10Be concentrations in spent resins were measured to be about 10 ng/g, 20 pg/g and 4 ng/g of dry resin, respectively. Considering 129I and 41Ca, the measured concentrations agree with those assessed from scaling factors established relatively to easily measured gamma emitters (137Cs and 60Co). For 10Be, the presented results are significantly different from expected values but are in agreement with previous ICP-MS results.