This paper studies the nuclear technology and evaluates the most likely technical developments to come until 2030. It examines the features of Czech and French nuclear programs, compares the structures of nuclear sectors and estimates the advantages & drawbacks of further developments in both countries. The paper assesses if certain developments of the nuclear sector in one country can be used in the other country and vice-versa. Finally, proposes a set of recommendations for both countries regarding the development of their nuclear program.

Le démantèlement des réacteurs Uranium Naturel Graphite Gaz génèrera en France environ 23 000 tonnes de déchets radioactifs graphités. La gestion appropriée de ces déchets nécessite de déterminer leur inventaire radiologique et de disposer de données fiables sur la localisation et la spéciation des radionucléides (RN). Le 14C a été identifié comme RN d'intérêt pour le stockage en raison de son inventaire initial important et du risque de présence d'une fraction organique mobile dans l'environnement, lors de la phase de stockage. A ce titre, l'objectif de cette thèse CIFRE, réalisée en partenariat avec EDF, est de mettre en œuvre des études expérimentales permettant de simuler et d'évaluer l'impact de la température, de l'irradiation et de la corrosion radiolytique du graphite sur le comportement migratoire en réacteur du 14C et de son précurseur azote. Les données ainsi acquises sont intégrées dans la deuxième partie de ce travail consacrée à l'étude d'un procédé de décontamination thermique du graphite en présence de vapeur d'eau. La démarche expérimentale consiste à simuler respectivement la présence de 14C et de 14N par implantation ionique de 13C et d'azote (14N ou 15N) dans un graphite de rondin SLA2 vierge. Cette étude montre que dans la gamme de températures du graphite en réacteur (100 - 500°C) et en absence de corrosion radiolytique, le 13C est stable thermiquement quel que soit l'état de structure du graphite. En revanche, les expériences d'irradiation du graphite chauffé à 500°C au contact d'un gaz représentatif du caloporteur radiolysé montrent le rôle synergique joué par les espèces oxydantes et l'endommagement du graphite favorisant la mobilité du 13C par gazéification des surfaces et/ou oxydation sélective du 13C plus faiblement lié. En ce qui concerne l'azote constitutif, il a tout d'abord été démontré que sa concentration en surface atteint plusieurs centaines de ppm (< 500 ppm at.) et décroît en profondeur jusqu'à environ 160 ppm at.. Contrairement au 13C implanté, l'azote implanté migre à 500°C lorsque le graphite est fortement déstructuré (environ 8 dpa) alors qu'il reste stable pour un taux de déstructuration moindre (0,14 dpa). Les expériences montrent également le rôle synergique des excitations électroniques et de la température qui accélèrent le transport de l'azote vers la surface du graphite. Cette migration de l'azote semble se faire sous forme moléculaire d'espèces C-N, C=N voire C N. Après huit heures d'irradiation ces espèces ne sont toutefois pas ou peu relâchées et restent bloquées à la surface. L'étude du procédé de décontamination thermique en présence de vapeur d'eau a nécessité la mise en place d'un dispositif de thermogravimétrie couplé à un générateur de vapeur d'eau ainsi que l'optimisation des paramètres de l'étude. Les influences de la température (700°C et 900°C) et de l'humidité relative (50 % HR et 90 % HR) ont été testées à un débit de gaz humide fixe de 50 mL/min. Dans ces conditions, l'oxydation sélective du carbone implanté a été confirmée / The dismantling of UNGG reactors in France will generate about 23 000 tons of radioactive graphite wastes. To manage these wastes, the radiological inventory and data on radionuclides (RN) location and speciation should be determined. 14C was identified as an important RN for disposal due to its high initial activity and the risk of release of a mobile organic fraction in environment, after water ingress into the disposal. Hence, the objective of this thesis, carried out in partnership with EDF, is to implement experimental studies to simulate and evaluate the impact of temperature, irradiation and graphite radiolytic corrosion on the in reactor behavior of 14C and its precursor, 14N. The obtained data are then used to study the thermal decontamination of graphite in presence of water vapor. The experimental approach aims at simulating the presence of 14C and 14N by the respective ion implantation of 13C and 14N or 15N in virgin graphite. This study shows that, in the temperature range reached during reactor operation, (100-500°C) and without radiolytic corrosion, 13C is thermally stable whatever the initial graphite structure. Moreover, irradiation experiments were performed on heated graphite (500°C) put in contact with a gas representative of the radiolysed coolant gas. They show the synergistic role played by the oxidative species and the graphite structure disorder on the enhancement of 13C mobility resulting in the gasification of the graphite surface and/or the selective oxidation of 13C more weakly bound than 12C. Concerning the pristine nitrogen, we showed first that the surface concentration reaches several hundred ppm (<500 ppm at) and decreases at deeper depths to about 160 ppm at.. Unlike implanted 13C, implanted nitrogen migrates at 500 ° C when the graphite is highly disordered (about 8 dpa) while remaining stable for a lower disorder rate (0.14 dpa). Experiments also show the synergistic role by electronic excitations and temperature that accelerate the transport of nitrogen to the surface of the graphite. Nitrogen seems to migrate in the form of molecular species (CN, C = N or C N). After eight hours of irradiation these species are, however, little or not released and blocked at the surface. The study of the thermal decontamination of graphite in presence of water vapor was performed with a thermogravimetric device coupled to a steam water generator device. The influence of temperature (700 ° C and 900 ° C) and of the relative humidity (50% RH and 90% RH) was tested with a wet gas fixed flow rate of 50 ml/min. Under these conditions, the selective oxidation of implanted carbon was confirmed

Ce travail se situe dans le cadre des études sur la gestion des déchets graphites des centrales nucléaires Uranium Naturel Graphite Gaz (UNGG) de première génération. Leur fonctionnement a généré 23000 tonnes de déchets graphites pour lesquels la loi du 28 juin 2006 prévoit un stockage dédié. La gestion à long terme de ces déchets nécessite de prendre en compte deux radionucléides principaux : le ^14C et le ^36Cl, principaux contributeurs de dose sur le long terme. Afin de consolider les données sur l'inventaire de ces radionucléides et de prévoir leur comportement lors de la resaturation en eau du site de stockage, il est nécessaire de disposer de données liées à leur distribution et à leur spéciation dans le graphite avant stockage. Ce travail a été centré sur l'étude du chlore. Il a eu pour objectif de retracer le comportement du 36Cl dans le graphite nucléaire durant “sa vie” en réacteur et, en particulier d'étudier les effets de la température et de la corrosion radiolytique de manière découplée. Nos résultats permettent de déduire qu'il se produit un relâchement rapide du 36Cl d'environ 20% dès les premières heures de fonctionnement du réacteur. Celui-ci est suivi par un relâchement beaucoup plus lent tout au long de la vie du réacteur. Nous avons identifié la présence de deux fractions distinctes de chlore correspondant à des formes chimiques différentes (n'ayant pas la même stabilité thermique) ou à deux localisations du chlore d'accessibilités différentes. Notre etude montre également que la corrosion radiolytique semble promouvoir le relâchement du chlore et cela quelle que soit la dose d'irradiation. La forme chimique du chlore est majoritairement organique. / This work concerns the dismantling of the UNGG reactor which have produced around 23 000 t of graphite wastes that ave to be disposed of according to the Frenche law of June 206. These wastes contain two long-lived radionuclides (^ 14C and ^36Cl) which are the main long term dose contributors. In order to get information about their inventory and their long term behaviour in case of water ingress into the repository, it is necessary to determine their location and speciation in the irradiated graphite after the reactor shutdown. This work concerns the study of ^36Cl. The main objective is to reproduce its behaviour during reactor operation. For that purpose, we have studied the effects of temperature and radiolytic corrosion indepently. Our results show a rapid release of around 20% ^36Cl during the first hours of reactor operation whereas a much slower release occurs afterwards. We have put in evidence two types of chlorine corresponding to two different chemical forms (of different thermal stabilities) or to two locations (of different accessibilities). We have also shown that the radiolytic corrosion seems to enhance chlorine release, whatever the irradiation dose. Moreover, the major chemical form of chlorine is inorganic.

Sedimentary rocks are characterized with very low permeability (in the order of 10-22 m2), low diffusivity, a possible self-healing of fractures, and a good capacity to retard radionuclide transport. In recent years, sedimentary rocks are investigated by many research groups for their suitability for the disposal of radioactive waste. Development of deep geologic repositories (DGRs) for the storage of radioactive waste within these formations causes progressive modification to the state of stress, to the groundwater regime, and to the chemistry of the rock mass. Thermal effects due to the ongoing nuclear activity can cause additional disturbances to the system. All these changes in the system are coupled and time-dependent processes. These coupled processes can result in the development of an excavation damaged zone (EDZ) around excavations. More permeable than the undisturbed rock, the EDZ is likely to be a preferential pathway for water and gas flow. Consequently, the EDZ could be a potential exit pathway for the radioactive waste to biosphere. An investigation of the Hydraulic-Mechanical (HM) and Thermal-Hydraulic-Mechanical-Chemical (THMC) behaviour of sedimentary rock formations is essential for the development of DGRs within such formations. This research work consists of (1) an experimental investigation of the mechanical behaviour of the anisotropic Tournemire argillite, (2) modeling of the mechanical behaviour of the Tournemire argillite, and (3) numerical simulations of the mechanical and hydromechanical behavior of two host sedimentary rocks, the Tournemire argillite and Cobourg limestone, for deep geological repository for nuclear wastes. The experimental program includes the measurements of the physical properties of the Tournemire argillite and its mechanical response to loading during uniaxial compression tests, triaxial compression tests with different confining pressures, unconfined and confined cyclic compression tests, Brazilian tests, and creep tests. Also, acoustic emission events are recorded to detect the initiation and propagation of microcracks within the rock during the uniaxial testing. The approach for modeling the mechanical behaviour of the Tournemire argillite consists of four components: elastic properties of the argillite, a damage model, the proposed concept of mobilized strength parameters, and the classical theory of elastoplasticity. The combination of the four components results in an elastoplastic-damage model for describing the mechanical behaviour of the Tournemire argillite. The capabilities of the model are evaluated by simulating laboratory experiments. Numerical simulations consist of: (1) a numerical simulation of a mine-by-test experiment at the Tournemire site (France), and (2) numerical simulations of the mechanical and hydromechanical behaviour of the Cobourg limestone within the EDZ (Canada). The parameters influencing the initiation and evolution of EDZ over time in sedimentary rocks are discussed.

Sedimentary rocks are characterized with very low permeability (in the order of 10-22 m2), low diffusivity, a possible self-healing of fractures, and a good capacity to retard radionuclide transport. In recent years, sedimentary rocks are investigated by many research groups for their suitability for the disposal of radioactive waste. Development of deep geologic repositories (DGRs) for the storage of radioactive waste within these formations causes progressive modification to the state of stress, to the groundwater regime, and to the chemistry of the rock mass. Thermal effects due to the ongoing nuclear activity can cause additional disturbances to the system. All these changes in the system are coupled and time-dependent processes. These coupled processes can result in the development of an excavation damaged zone (EDZ) around excavations. More permeable than the undisturbed rock, the EDZ is likely to be a preferential pathway for water and gas flow. Consequently, the EDZ could be a potential exit pathway for the radioactive waste to biosphere. An investigation of the Hydraulic-Mechanical (HM) and Thermal-Hydraulic-Mechanical-Chemical (THMC) behaviour of sedimentary rock formations is essential for the development of DGRs within such formations. This research work consists of (1) an experimental investigation of the mechanical behaviour of the anisotropic Tournemire argillite, (2) modeling of the mechanical behaviour of the Tournemire argillite, and (3) numerical simulations of the mechanical and hydromechanical behavior of two host sedimentary rocks, the Tournemire argillite and Cobourg limestone, for deep geological repository for nuclear wastes. The experimental program includes the measurements of the physical properties of the Tournemire argillite and its mechanical response to loading during uniaxial compression tests, triaxial compression tests with different confining pressures, unconfined and confined cyclic compression tests, Brazilian tests, and creep tests. Also, acoustic emission events are recorded to detect the initiation and propagation of microcracks within the rock during the uniaxial testing. The approach for modeling the mechanical behaviour of the Tournemire argillite consists of four components: elastic properties of the argillite, a damage model, the proposed concept of mobilized strength parameters, and the classical theory of elastoplasticity. The combination of the four components results in an elastoplastic-damage model for describing the mechanical behaviour of the Tournemire argillite. The capabilities of the model are evaluated by simulating laboratory experiments. Numerical simulations consist of: (1) a numerical simulation of a mine-by-test experiment at the Tournemire site (France), and (2) numerical simulations of the mechanical and hydromechanical behaviour of the Cobourg limestone within the EDZ (Canada). The parameters influencing the initiation and evolution of EDZ over time in sedimentary rocks are discussed.

Les Argiles de Boom ont été identifiées par le SCK-CEN comme un éventuel site de stockage de déchets nucléaires en couche géologique profonde : elles font l’objet d’études dans le laboratoire souterrain de Mol (Belgique). Dans ce contexte, il est important d’évaluer les conséquences du creusement de galeries sur les propriétés de ces Argiles. Ce travail de thèse cible plus particulièrement les effets d’oxydation à l’air et de biodégradation anaérobie sur la MO fossile. Les expériences d’oxydation ont combiné des suivis expérimentaux (série artificielle) et des études sur des échantillons altérés in situ (série naturelle) dans les galeries du laboratoire. Elles ont ciblé le kérogène, la MOE et MOD. La confrontation des données des deux séries révèle que nos simulations expérimentales sont représentatives des processus ayant lieu dans les galeries. Ces travaux démontrent aussi que l’oxydation induit (i) une augmentation importante de la quantité de MOD et (ii) une modification majeure de la chimie de la MOE et de la MOD caractérisée par la formation de molécules oxygénées de faible poids moléculaire. Par ailleurs, l’étude comparative des eaux issues des échantillons altérés avec celles prélevées dans les piézomètres du site démontre que ces dernières ne sont pas affectées par des processus d’oxydation et sont comparables aux eaux issues des échantillons sains. Des expériences de biodégradation menées sur des argiles saines et préalablement oxydées artificiellement n’ont montré aucune évolution significative de la MO fossile (MOE et MOD); la biodégradation anaérobie n’est donc pas un processus dominant dans le contexte des perturbations induites par les excavations / The Boom Clay was focused because it is identified by SCK-CEN as a possible radioactive waste storage in the geological disposal site and in situ experiments are performed in the underground laboratory of Mol (Belgium). In this context, it is important to assess the consequences of galleries excavation on the properties of the Boom Clay. The particular focus of this study is the effects of air oxidation and anaerobic biodegradation on the OM. The experiments dealing with the effects of air oxidation have combined studies of artificial oxidized samples (artificial series) and samples altered in the gallery of the underground laboratory (natural series). These experiments focus on the Kerogen, the EOM and the DOM. The comparison of data from artificial and natural series shows firstly that our experimental simulations are the representative of processes taking place in the galleries. These studies show that air oxidation induced (i) a significant increase in the amount of DOM and (ii) a major change in the chemistry of the EOM and DOM characterized by the formation of low molecular weight oxygenated molecules. Moreover, comparison between water extracted from altered samples and piezometers shows that the water of the site is not affected by oxidation processes. The piezometer water samples are similar to that extracted from non-altered samples. Biodegradation experiments conducted on non altered and artificially oxidized clay did not show significant changes of fossil and dissolved organic matter. It seems that anaerobic biodegradation is not a major process in the context of disturbances induced by the excavation

It is a generally agreed internationally that high level nuclear wastes containing long-lived radioactive wastes should be disposed in deep and stable geological formations that are 500-1000 m below ground level. Deep geological disposal is based on the concept of multiple barriers to prevent deep ground-waters, present in almost all rock formations, from rapidly leaching the wastes and transporting radioactivity away from the repository. The multiple barrier system comprises of ‘engineered barriers’ that are constructed in the repository and ‘natural barriers’ in the surrounding geological environment. The engineered barrier components comprise of the vitrified solid waste, canister (to contain the vitrified waste), and a buffer or backfill material (clay or cement) that fills the annular space between the canister and the walls of the hole drilled in the floor of host-rock. The natural barrier is provided by the rocks and soils between the repository and earth’s surface. The canisters containing the hig level waste (HLW) upon placement in DGR need protection against tectonic activities and chemical attack by dissolved elements and from microbes. Densely compacted bentonite is identified suitable for this purpose owing to its large swell potential, low permeability, sufficient bearing capacity and high cation adsorption capacity. In the deep geological repository (DGR) for disposal of high level nuclear wastes, iodine-129 is one of the significant nuclides, owing to its long half-life (half life = 16 million years) and tendency to easily migrate out of the geological repository into the biosphere caused by its high solubility and poor sorption onto most geologic media. Bentonite buffer by virtue of negatively charged basal surface has negligible affinity for retention of iodide anions. Attempts have been made to improve the iodide retention capacity of bentonite by treating the clay with cationic polymers, this however occurs at the cost of reduced swelling ability of bentonite clay. The compacted bentonite employed in deep geological repositories must possess large swell potential to enable it to close fissures and cracks that form on drying of the expansive clay by the heat arising from the high level nuclear waste and thereby close pathways for migration of radionuclides (from breached canister) to the geo-environment. Therefore, it becomes important to identify an additive that enhances the iodide retention ability of the mix without significantly impairing its swelling ability. Based on the strong affinity of silver for iodide ions, the feasibility of mixing silver-kaolinite (termed AgK) clay with bentonite to improve the latter’s iodide sorption capacity and the impact of mixing AgK clay with bentonite on swelling ability of the mix forms one of the the focus of this thesis. Silver-kaolinite clay was prepared by heating 80% kaolinite + 20% silver nitrate mix at 400°C for 30 min, followed by washing (to remove unreacted silver nitrate) and oven-drying the resultant AgK clay. Physical mixing of AgK and bentonite was considered a viable proposition as small additions (10% to 20% on dry mass basis) besides imparting iodide sorption ability was expected to have minor influence on the swelling ability of the mix. As organo-bentonites are known to retain iodide ions, it was considered relevant to compare the iodide removal behaviour of AgK and organo¬bentonite clay. Hexadecylpyridinium-bentonite (termed as HDPy+B) is the organo¬bentonite examined in this thesis and is prepared by treating bentonite with hexadecylpyridinium chloride mono hydrate salt (C21H38ClN.H2O; molecular weight = 358.01). The hexadecylpyridinium chloride mono hydrate salt is a cationic quaternary ammonium compound and has been used by earlier researchers to prepare organo-bentonite for removal of iodide ions from aqueous solutions. The impact of mixing AgK and HDPy+B clays on the iodide retention and swelling behaviour of bentonite is also considered in the thesis. The mass-balance calculations, XRD analysis, X-ray photon emission survey spectrum and EPMA tests performed on kaolinite-silver nitrate mix/AgK/kaolinite specimen indicated that silver occurs as uniform coatings of AgO/Ag2O on kaolinite surface of the AgK specimen. The AgK clay has strong affinity for iodide ions reflected by the large distribution coefficients (Kd) values of 1367 and 293 mL/g at initial iodide concentrations of 750 mg/L and 1000 mg/L. Further, the sorption process was rapid, unaffected by the presence of co-ions, elevated temperature of sorption and was practically irreversible at range of pH conditions. The iodide retention by AgK is attributed to occurrence of hydrolysis and exchange reactions. On contacting the AgK with water, the AgO species hydrolyze to form AgOH; iodide ions are retained by replacing the hydroxyl group of AgOH leading to formation of AgI phase. The adsorption of HDPy+Cl- ions by bentonite occurs by replacement of the native exchangeable cations by HDPy+ ions and adsorption by van der Waals interactions between the organic cations and the clay surface. The adsorbed cationic polymer neutralize the negative charge of the clay surface. Zeta potential measurements of HDPy+B specimen indicated that adsorption of cationic polymer transforms the negatively charged clay particles into positively charged particles that favour anion adsorption. Sorption of iodide ions by HDPy+B specimen exhibits two distinct segments: 1) the iodide sorption increased rapidly at lower iodide concentration (91 mg/L to 475 mg/L) and are retained by Coulombic adsorption to the cationic groups contained in the loops and tails of the adsorbed polymer (primary adsorption sites) and 2) the relatively slower adsorption at higher iodide concentrations (larger than 475 mg/L) is attributed to exchange with chloride ions attached to HDPy+Cl-ion pair (secondary adsorption sites). The Kd values for iodide adsorption vary from 15 mL/g to 184 mL/g at initial iodide concentrations of 91 mg/L to 996 mg/L respectively. Comparing the iodide removal efficiencies of AgK and HDPy+B specimens revealed that the AgK clay exhibited larger iodide removal; further while the iodide removal by AgK specimen was almost instantaneous (complete in < 5 min), iodide removal by HDPy+B specimen was a slow process (18-24 h is needed to attain equilibrium). Likewise, the iodide retention capacity of the 50%B-50%HDPy+B mix (B = bentonite) is substantially smaller than of the 90%B-10%AgK and 80%B¬20%AgK mixes. Cation exchange capacity (CEC) measurements brought out that mixing AgK with bentonite besides imparting an iodide retention capacity essentially retains the large cation exchange capacity of the expansive clay. On the other hand mixing HDPy+B with bentonite imparts a smaller iodide retention capacity to the mix and leads to a notable reduction in the CEC of the expansive clay. Results of oedometer swell tests brought out that dilution of bentonite with 10% and 20% AgK specimen does not impact its swell potential and leads to some (10%) reduction in swell pressure, while dilution with 50% HDPy+B clay leads to notable (58%) reduction in swell potential and swell pressure (21%) underlining the superiority of AgK specimen as additive to bentonite in deep geological repositories. The swell pressure of the compacted 50%B-50%HDPy+B mix is 21% lower than that of the compacted bentonite specimen. Comparatively, dilution of bentonite with 10% and 20% AgK specimen induces 8-10% lower swell pressure in comparison to the undiluted counterpart. Swell pressure results of compacted 80%B-20%HDPy+B mix is not considered as this mix was unable to retain iodide ions. Superposing the field 129I concentration levels on I removal efficiency indicate that use of 90%B-10%AgK mix would suffice to provide 100% iodide removal efficiency and ensure that the swelling characteristics of bentonite is least affected by dilution.