The aim of diploma thesis is to analyse the overall technical base and to monitor logistic processes which are from the very beginning associated with radioactive waste and spent nuclear fuel. The main emphasis is put on spent nuclear fuel from nuclear power plants. The current situation in this area is illustrated by an example of the Czech Republic and Sweden. The comparison of Czech and Swedish system leads to demonstration of strengths and weaknesses of the used methods.

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.

Deep geological repositories (DGR) are considered an effective long-term solution for radioactive waste disposal. Sedimentary (argillaceous formations) and crystalline rocks are currently under investigation worldwide as potential host formations for DGR. Their low porosity (<1-2 %) and very low hydraulic conductivity result in diffusion-dominated solute transport. There is a need to investigate their diffusion properties in detail, the long-established diffusion methods do not allow an evaluation of the spatial relationship between tracers and the characteristics of the geological medium. The aim of this project was to measure diffusion coefficients in low-porosity rocks (< 2%) using X-ray radiography and iodide tracer. The method is a non-destructive technique based on the principle of X-ray attenuation; it provides temporal- and spatially-resolved information of a highly attenuating tracer diffusing in a sample. Samples from the Cobourg Formation, an Ordovician argillaceous limestone from the Michigan Basin, and from the Lac du Bonnet batholith, an Archean granitic pluton were used in this study. X-ray radiography data from the Cobourg Formation indicate tracer accumulation occurs on dark argillaceous layers in the rock characterized by clay minerals and organic matter. It is proposed that the I– tracer solution underwent photo-chemical oxidation, leading to the formation of I2, a highly reactive volatile iodine species and I3–, which readily reacted with humic substances contained in the clay- and organic rich zones in the limestone samples. In the case of the granitic samples, attempts at measuring diffusion coefficients encountered several challenges. The results indicate that tracer signal can be detected, however diffusion signal is masked by imaging errors and noise.

The diploma thesis was elaborated on the topic of public awareness on issues related to the nuclear waste repository. The issue of nuclear waste management is currently being updated and increasingly discussed. Especially in connection with the search for a new site for the construction of a deep repository of nuclear waste and spent nuclear fuel. The aim of the thesis was to determine the level of knowledge of the population in the field of nuclear waste, its management, knowledge of the current nuclear waste repositories and the intended deep repository of nuclear waste and spent nuclear fuel and, last but not least, of ionizing radiation. The other task was to compare the level of knowledge of the inhabitants living in one of the sites of the intended underground repository (the Čihadlo site) and the inhabitants living outside this site. The following hypotheses have been established: "The level of knowledge on issues related to the repository for nuclear waste and spent nuclear fuel will be statistically significantly higher for residents living in the Čihadlo site than for those living outside that site" and "Knowledge on issues related to the nuclear waste repository and of spent nuclear fuel will reach at least 70% in both groups." A questionnaire survey was conducted to achieve the objectives set and verify the hypotheses. The results were evaluated using descriptive and mathematical statistics. The questionnaire consisted of 20 questions and 100 respondents from each site. The hypothesis has been confirmed that the level of knowledge of the inhabitants living in the Čihadlo site is statistically significantly higher. Knowledge on issues related to the repository for nuclear waste and spent nuclear fuel reached at least 70% for all respondents only on some issues. In the diploma thesis, there was a picture of the level of knowledge of the inhabitants about nuclear waste, its handling and storage of nuclear waste. The results obtained could be used as one of the bases in the site selection process for building a deep repository.

The master's thesis deals with the analysis of source term of the spent nuclear fuel of the Dukovany Nuclear Power Plant in order to determine the proposals for the transfer of spent nuclear fuel to a deep geological repository in the Czech Republic. To introduce the reader into the issue are briefly described the main aspects, such as the development of the nuclear fuel used in the history of the Dukovany Nuclear Power Plant. These aspects have an influence on the final draft of the timetable. One of the important partial tasks is the processing of an estimate of the future range of spent nuclear fuel, which is based on the current ideas of company ČEZ, a.s. for the future direction of the fuel cycle at the Dukovany nuclear power plant. For the purposes of this work, the key data are the time dependencies of radioactivity and the development of residual heat in individual fuel assemblies. This data are calculated by the software called PAL440_R4, based on the prepared estimates of the spent nuclear fuel assortment. The calculated data are then edited and sorted by MS Excel. For the sake of completeness, the characteristic values and the time dependencies of the radioactivity and the development of residual heat in fuel assemblies. Final timetables for the transfer of spent nuclear fuel to a deep geological repository are processed in several variants, and their selection and application options are justified. For illustration are important parameters given in the form of tables and charts.

The thesis titled Legal aspects of constructing a deep geological repository od nuclear waste in the Czech Republic deals with the administrative procedures that needs to be taken before beginning construction of a deep geological repository. Work can be divided into three major parts. The first part deals with analysis of current legislation relating to nuclear energy, with emphasis on the treatment of nuclear waste. International and European commitments of the Czech Republic, current and new Atomic act, as well as institutional and financial arrangements for nuclear waste management are also included in this part. The following section has been devoted to the various administrative procedures. The goal of this section in not to provide complete description of the procedures, but to highlight points of interest and identify potential problems of current legislation and to propose better solution. The last major part is dedicated to public participation in the various administrative procedures. Emphasis is placed on the possibility of involvement of public and the affected communities in related administrative procedures. This chapter also contains a draft of the bill on community involvement in the process of selecting a site for deep repository.

Canada produces about one-third of the global supply of medical radioisotopes. The nuclear power reactors in Ontario, Quebec and New Brunswick have generated about 17 percent of the electricity in the country every year (NWMO, 2010; Noorden; 2013). Since the 1960s, more than 2 million used (or spent) fuel bundles (high-level radioactivity) and 75,000 m³ of low- and intermediate-level radioactive waste have been produced, which is increasing by 2000 to 3000 m³ every year after reducing the processed volume (Jensen et al., 2009). More than 30 countries around the world, including Canada, have proposed construction of very deep geological repositories (DGRs) to store this nuclear waste for design periods 1,000,000 years. DGR concepts under development in Canada (the DGR is likely to be constructed in Ontario) are based on a multi-barrier system (NWMO, 2012). A crucial component of the multi-barrier system is the engineered barrier system (EBS), which includes a buffer, backfill, and tunnel sealing materials to physically, chemically, hydraulically and biologically isolate the nuclear waste. Bentonite-based material has been chosen for this critical use because of its high swelling capacity, low hydraulic conductivity, and for its good ability to retain radionuclides in the case of failed canisters. However, the presence of bentonite-based material in DGRs, surrounded by an aggressive environment of underground saline water, nuclear waste heat decay, and corrosion products under confining stress, may lead to mineralogical changes. Consequently, the physical and physiochemical properties of bentonite-based materials may change, which could influence the performance of bentonite in an EBS as well as the overall safety of DGRs. The objective of this research is to investigate the impact of the underground water salinity, heat generated by nuclear waste, and corrosion products of nuclear waste containers in Ontario on the engineering and micro-/nano-structural properties of bentonite-sand engineered barrier materials. Free-swelling, swelling pressure and hydraulic conductivity tests have been performed on bentonite-sand mixtures subjected to various chemical (groundwater chemistry; corrosion water with iron as a corrosion product) and thermal (heat generated) conditions. Several techniques of micro- and nano-structural analyses, such as x-ray diffraction (XRD), X-Ray microanalysis (DES), surface area and pore size distribution analyses (BET, BJH) and differential gravimetric (TGA and DTG) analyses have also been conducted on the bentonite-sand materials. Valuable results have been obtained for better understanding the durability and performance of the bentonite-sand barrier for the DGR which may be located in Ontario. The obtained results have shown that the groundwater chemistry and corrosion products of the nuclear containers significantly deteriorate the swelling and permeability properties of the tested bentonite-sand barrier materials, while temperature has little or no effect.

Electricity generation at nuclear power plants produces a large amount of high-level radioactive waste (HLW) every year, which has long-term detrimental effects on humans and the environment. Other applications of nuclear technology (e.g., medicine, research, nuclear weapons, industry) also produce radioactive waste (e.g., low-level radioactive waste, LLW, Intermediate-level waste, ILW). The potential of deep geological repositories (DGRs) as an option for disposal of radioactive waste (HLW, ILW, LLW) has been examined in several countries, including Bulgaria, Canada, China, Finland, France, Germany, India, Japan, Russia, Spain, Sweden, Switzerland, Ukraine and the United Kingdom and are still under discussion. In Ontario, Canada, DGRs with a multi-barrier system comprised of a sedimentary rock formation (i.e., a natural barrier) and an engineered barrier system (EBS) are currently under consideration. An EBS consists of various components, such as waste containers, buffer, backfill, and tunnel sealing materials, intended to prevent the release of radionuclides. Several engineered barrier materials, including a mixture of bentonite and sand, are currently being considered for use in DGRs for nuclear waste in Ontario. Bentonite has some advantageous physical and chemical properties, such as low permeability, high plasticity, and high swelling potential, which provide it with a good sealing ability and thus make it an effective barrier. However, interaction between the compacted bentonite–sand mixture and underground water chemistry fluids (chemical factor) in the DGR could significantly alter the favourable properties of bentonite (e.g., swelling potential), thus influencing its performance when used in an EBS and eventually jeopardizing the overall safety of DGRs. In addition, other parameters, such as the clay content, initial dry density and moisture content of the compacted barrier (physical factors), as well as the presence of salts in groundwater may affect the physical and physiochemical properties of barrier materials. Moreover, during the lifetime of a DGR for used spent fuel, the bentonite–based barrier material will not only be exposed to a broad range of groundwaters with different chemical compositions, but also to high temperatures (heat generated by the nuclear wastes) (thermal factor). Thus, the interaction between the compacted bentonite–sand mixture, the surrounding groundwater and the heat from the nuclear waste material could jeopardize the favourable properties of the bentonite-based (bentonite-sand) barrier material. Properties of a bentonite-sand barrier is an important characteristic to study while designing and constructing an EBS for a DGR. Thus, to understand and assess the operations of DGRs in Ontario, comprehensive studies must be performed on engineering properties like swelling behaviour, permeability, and thermal conductivity. The goal of this research study is to experimentally investigate the physical, chemical and thermal factors that influencing the engineering properties of a barrier material made up of bentonite-sand composite used in DGRs for nuclear waste in Ontario. Compacted samples are subjected to one-dimensional free swell test to understand the swelling behaviour of the material. Hydraulic conductivity was investigated using a flexible wall permeability test. Thermal conductivity and diffusivity were tested using Decangon KD2 Pro with TR-1 and and KS-1 sensors. The specimens contain different bentonite–sand mixture ratios (20:80, 30:70, 50:50, and 70:30 dry mass), and they are tested under conditions with differing bentonite content, dry density, groundwater chemistry, and temperature. Additional tests were conducted to investigate the microstructure of the specimens. These tests include X-ray diffraction (XRD) analysis, mercury intrusion porosimetry (MIP), and thermogravimetric analyses (TG/DTG). The results reveal that the time and strain required to achieve maximum swelling of compacted bentonite–sand specimens increase with the increase of initial dry density. The simulated saline solutions of Guelph and Trenton groundwater are found to suppress the swelling of the bentonite–sand specimens. This in turn leads to the increase of hydraulic conductivity and decrease of thermal properties of the barrier material. However, the impact of the salinity is significantly reduced by increasing the dry densities and sand content of the compacted material. Moreover, the coupled effect of salinity and temperature decreases the swelling potential of the bentonite-sand mixture. Also, some transformation of Na-montmorillonite into Ca-Montmorillonite was observed. The results also indicate that some montmorillonites might have been transformed into illites, thereby further decreasing the swelling potential of the bentonite-based barrier.

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.

Actuellement, bien que la plupart des pays nucléaires converge vers la même solution technique: le stockage profond pour la gestion des déchets radioactifs de haute activité et à vie longue, les objectifs calendaires divergent d'un pays à l'autre. Grâce au calcul économique, nous souhaitons apporter des éléments de réponse à la question suivante : En termes de temporalité, comment les générations présentes, qui bénéficient de la production d'électricité nucléaire, doivent-elles supporter les charges de la gestion des déchets radioactifs en tenant compte des générations futures ? Cette thèse se propose d'analyser spécifiquement la décision française en tenant compte de son contexte. Nous proposons un ensemble d'outils qui permet d'évaluer l'Utilité du projet de stockage profond en fonction des choix de temporalité. Notre thèse étudie également l'influence en retour des choix de stockage sur le cycle du combustible nucléaire. Au-delà, nous prenons en compte les interactions entre le stockage profond et les choix de parc nucléaire et de cycle du combustible qui constituent un « système complet ». / Nowadays, the deep geological repository is generally considered as the reference solution for the definitive management of spent nuclear fuel/high-level waste, but different countries have decided different disposal deployment schedules. Via the economic calculation, we hope to offer some answers to the following question: In terms of disposal time management, how should the present generations, benefiting from the nuclear power generation, bear the costs of radioactive waste management, while taking into account future generations? This thesis proposes to analyze specifically the French decision in its context. We propose a set of tools to evaluate the Utility of the deep geological repository project according to the deployment schedule choices. Our thesis also studies the influence of disposal choices on the nuclear fuel cycle. Beyond, we also take into account the interactions between the deep geological repository, nuclear fleet and cycle choices which constitute a "complete system".