The thesis deals with nuclear waste deep repositories in the law. With exception of long- term storage the nuclear waste deep repositories represent the only technical solution currently available to deal with the increasing volume of highly radioactive waste and spent nuclear fuel. The planning and construction of nuclear waste deep repository entails number of problems and challenges, both from technical and legal perspectives. The thesis aims to describe current legislation regarding the management of radioactive waste and spent nuclear fuel within the territory of the Czech Republic with a particular focus on planning and construction of a nuclear waste deep repository. There is step by step described procedure of planning and constructing a nuclear waste deep repository under the current legislation. Within this description the author evaluates the current legislation and identifies its major issues. Subsequently, the author reflects on the cause of these issues and proposes their solutions.

The closure of the nuclear fuel cycle is a topic of interest in the sustainability context of nuclear energy. The implication of such closure includes considerations of nuclear waste management. This originates from the fact that a closed fuel cycle requires recycling of useful materials from spent nuclear fuel and discarding of non-usable streams of the spent fuel, which are predominantly the fission products. The fission products represent the near-term concerns associated with final geological repositories for the waste stream. Long-lived fission products also contribute to the long-term concerns associated with such repository. In addition, an ultimately closed nuclear fuel cycle in which all actinides from spent nuclear fuels are incinerated will result in fission products being the only source of radiotoxicity. Hence, it is desired to develop a transmutation strategy that will achieve reduction in the inventory and radiological parameters of significant fission products within a reasonably short time. In this dissertation, a transmutation strategy involving the use of the VHTR is developed. A set of specialized metrics is developed and applied to evaluate performance characteristics. The transmutation strategy considers six major fission products: 90Sr, 93Zr, 99Tc, 129I, 135Cs and 137Cs. In this approach, the unique core features of VHTRs operating in equilibrium fuel cycle mode of 405 effective full power days are used for transmutation of the selected fission products. A 30 year irradiation period with 10 post-irradiation cooling is assumed. The strategy assumes no separation of each nuclide from its corresponding material stream in the VHTR fuel cycle. The optimum locations in the VHTR core cavity leading to maximized transmutation of each selected nuclides are determined. The fission product transmutation scenarios are simulated with MCNP and ORIGEN-S. The results indicate that the developed fission product transmutation strategy offers an excellent potential approach for the reduction of inventories and radiological parameters, particularly for long-lived fission products (93Zr, 99Tc, 129I and 135Cs). It has been determined that the in-core transmutation of relatively short-lived fission products (90Sr and 137Cs) has minimal advantage over a decay-only scenario for these nuclides. It is concluded that the developed strategy is a viable option for the reduction of radiotoxicity contributions of the selected fission products prior to their final disposal in a geological repository. Even in the cases where the transmutation advantage is minimal, it is deemed that the improvement gained, coupled with the virtual storage provided for the fission products during the irradiation period, makes the developed fission product transmutation strategy advantageous in the spent fuel management scenarios. Combined with the in-core incineration options for TRU, the developed transmutation strategy leads to potential achievability of engineering time scales in the comprehensive nuclear waste management.

The nuclides present in spent nuclear fuel are categorized according to their capacity for detection by Compton suppression or gamma-gamma coincidence methods. The fifty nuclides with the highest activities in spent fuel are identified, their decay schemes analyzed, and the best detection scheme for each is recommended. / text

La caractérisation isotopique des éléments présents dans les combustibles nucléaires irradiés est d'une importance majeure pour la qualification et la validation des codes de calculs neutroniques ainsi que la gestion des déchets nucléaires. Le protocole conventionnel pour l'analyse de ces échantillons nécessite plusieurs étapes de séparation par chromatographie liquide. L'uranium (U), le plutonium (Pu) et une fraction contenant les produits de fission et les actinides mineurs, sont séparés par chromatographie sur résine échangeuse d'ions puis les fractions purifiées d'U et de Pu sont analysées par spectrométrie de masse. L'objectif de cette thèse consiste à étudier et développer un protocole analytique applicable sur des échantillons de type combustibles irradiés et employant une technique séparative transposable sur plateforme miniaturisée qui devra pouvoir être couplée à un spectromètre de masse à source plasma et à système multicollection (ICPMS-MC) afin de réaliser en ligne l'analyse isotopique et élémentaire des éléments présents dans le combustible irradié. Une méthode de séparation de l'U, du thorium (Th) et des lanthanides par électrophorèse capillaire (EC) avec préconcentration de l'échantillon a été développée en utilisant le Th(IV) comme analogue chimique du Pu(IV). L'électrolyte de séparation se compose d'acide acétique 0,25M comme complexant ainsi que de sel d'ammonium 0,1M, pour ajuster la force ionique et permettre la préconcentration. Le montage d'EC a été adapté afin d'être intégré en boite à gants et couplé à un ICPMS-MC. La séparation de l'Am de l'U et du Pu a été réalisée sur quelques nL d'une solution de combustible irradié. / Precise isotopic and elemental characterization of nuclear spent fuel is a major concern for the validation of the neutronic calculation codes and waste management in the nuclear industry. The conventional protocol for the analysis of nuclear fuel samples uses several purification steps by liquid chromatography. Uranium (U) and plutonium (Pu) and a fraction containing fission products and minor actinides are separated using ion exchange chromatography prior to the isotopic characterization of the U and Pu fractions by multi-collector mass spectrometry techniques. The objective of the work presented is to develop a new analytical approach based on miniaturized separation techniques like capillary and microfluidic electrophoresis coupled with a multicollector inductively coupled plasma mass spectrometry (MC-ICPMS) detection for online isotopic ratio measurements. An electrophoretic separation method of U, Pu and fission products with a stacking step was developed using thorium (Th) as a chemical analog for Pu(IV). The separation electrolyte is composed of acetic acid (0.25M) as complexing agent for the separation and 0.1M of ammonium salt to realize the stacking step. The instrumentation was adapted to be used in glove box and directly coupled to a MC-ICPMS. The separation of Am, Pu and U was realized with few nL of a spent nuclear fuel solution. The reproducibilities obtained on the isotope ratios were in the order of few ‰ and comparable with those obtained with the conventional analytical protocol. This new protocol will help to reduce the quantities analyzed from µg to ng, the liquid waste volume scale from mL to µL and the sample volumes form µL to nL.

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 primary goal of this thesis was to commission an instrument for vibrational sum frequency spectroscopy (VSFS), and exploit it for the study of solid/gas interfaces; of ultimate interest is characterisation of substrate surfaces in humid environments. Such effort is motivated by interest in understanding the potential for atmospheric corrosion in dry storage facilities of spent nuclear fuels or other nuclear-related wastes. VSFS is a non-linear, interface specific, vibrational spectroscopy, in which two photons of different energies (infrared (IR) and visible (VIS)) impinge upon a surface at the same point at the same time, leading to the generation of a third (sum-frequency generation (SFG)) photon. Features in VSFS spectra can be assigned to vibrational modes of interfacial species, and so enable details of interfacial structure and chemistry to be elucidated. An instrument for such measurements has been developed using laser facilitates located in the Photon Science Institute (PSI) of The University of Manchester. More specifically, an ultra-fast (femtosecond) laser has been employed as a light source, enabling acquisition of spectra (~250 cm-1 in width at a resolution of ~11 cm-1) without the need for scanning the energy of either IR or VIS beams, i.e. so called broad-band VSFS. To test performance, data have been acquired from self-assembled monolayers of alkanethiols (octadecanethiol) on gold substrates, which demonstrate the utility of the instrument. Subsequent to commissioning, the VSFS instrument was initially exploited to study the interaction of two organic molecules, acetonitrile and acetic acid, with a single crystal TiO2(110) substrate; measurements were performed with the sample exposed to the vapour of each organic species under ambient conditions. Surface adsorption was identified through the appearance of the CH3 symmetric stretch. Furthermore, spectra as a function of light (IR/VIS/SFG) polarization combinations have been recorded to explore adsorbate angular geometry. Finally, VSFS measurements have been undertaken from a number of substrates (GaAs, Au, Zn, Fe, Cr, stainless steel), as a function of relative humidity; D2O was employed to overcome the issue of loss of IR beam intensity due to interaction with atmospheric H2O. Signal quality varies significantly with substrate, with the most insight being gained for the interaction of D2O with polycrystalline Zn. Clear vibrational resonances due to both hydroxyls (OD) and molecular water (D2O) are observed, which vary with relative humidity, indicating that there are significant changes in interface structure with relative humidity.

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.

A geological repository for spent nuclear fuel (SNF) and an associated encapsulation facility will be built in Sweden.  The encapsulation facility is planned to be in operation in 2025 and it will be the last place where verifying safeguards measurements of SNF can be performed. It is not clear what types of measurements that will be performed, because such requirements are not yet posed by national and international authorities and inspecting organizations. This report describes the objective and most recent results of a master thesis project, whereby a few existing non-destructive assay techniques for verifying SNF are selected for a review. The study focuses on the verifying ability of different techniques, or system of techniques in relation to the requirement that may be put on the future encapsulation plant. In addition, possible needs for future simulations and measurements are discussed. The work is done as a collaboration between Uppsala University in Sweden and Los Alamos National Laboratory in the USA.

Slutförvaring av använt kärnbränsle måste vara säker under 100 000 år eller mer för att förhindra att miljön skadas och att människor påverkas av långlivade radionuklider. Även om anläggningar för geologiskt djupförvar är utformade för att vara hållbara i många år, kan använt kärnbränsle komma i kontakt med grundvattnet i händelse av att flera barriärer brister. Det använda kärnbränslets inneboende radioaktivitet orsakar radiolys av inträngande grundvatten som producerar oxiderande och reducerande ämnen. Bland de radiolysprodukter som bildas rapporteras väteperoxid (H2O2) som en av de främsta orsakerna till oxidativ upplösning av bränslematrisen, UO2. Även om UO2 har låg löslighet i anoxiskt grundvatten, har oxiderad UO2, UO22+, flera storleksordningar högre löslighet. Detta utgör en risk för att radionukliderna släpps ut i miljön. Bikarbonat (HCO3-) är en av de viktigaste komponenterna i grundvatten och är känd för att öka upplösningen av UO22+. I denna studie undersöktes därför effekterna av HCO3- koncentrationen på den oxidativa upplösningen av UO2 genom att hålla den ursprungliga mängden H2O2 konstant på 0,2 mM och ändra HCO3- koncentrationen (1 mM, 2 mM, 5 mM och 10 mM). Dessutom undersöktes effekten av UO22+ på specieringen av H2O2 genom att tillsätta uranylnitrat (UO2(NO3)2 x 6H2O) till systemen innan de exponerades för H2O2. Specieringens inverkan på kinetiken för oxidativ upplösning av UO2 analyserades. Som ett resultat av experimenten har man dragit slutsatsen att mängden upplöst UO22+ är högre vid högre HCO3- koncentration. Dessutom minskar upplösningshastigheten för UO22+ med initial tillsats av UO22+ på grund av de komplex som bildas i systemen. Det observerades att oxidation av UO2 är den hastighetsbegränsande reaktionen i början av exponeringen, och att upplösningen av UO22+ därför fördröjs. Å andra sidan har man sett att bristen på HCO3- begränsar systemens upplösningsförmåga. Fri H2O2 är den dominerande formen av peroxid i systemen utan initialt tillsatt UO22+, medan -6 och -2 laddade komplex är dominerande i systemen med initialt tillsatt UO22+. H2O2-komplexen är mer effektiva på ytmekanismen i de system som har lägre HCO3- koncentration. Det finns ingen observerbar trend i H2O2-förbrukningshastigheten med avseende på HCO3-koncentrationen. Därför drogs slutsatsen att H2O2-förbrukningen är oberoende av upplösningsreaktionen. Slutligen följer upplösningen i systemet utan ursprungligt tillsatt UO22+ första ordningens kinetik med avseende på HCO3- koncentrationen. / Disposal of spent nuclear fuels is of great importance to prevent the environment and humans from being affected by long-lived radionuclides for 100,000 years or more. Even though the deep geological repositories are designed to remain durable for many years, spent nuclear fuel may come in contact with groundwater in case of a multi-barrier failure. The inherent radioactivity of spent nuclear fuel causes water radiolysis producing oxidizing and reducing agents. Among the radiolysis products, hydrogen peroxide (H2O2) is reported as a primary contributor to the oxidative dissolution of the fuel matrix, UO2. Although UO2 has low solubility in water, oxidized UO2, UO22+ , has several orders of magnitude higher solubility. This poses the risk of the radionuclides being released into the environment. Bicarbonate (HCO3–) is one of the main components of groundwater and is known to increase the dissolution of UO22+. Therefore, in this study, the effects of HCO3– concentration on the oxidative dissolution of UO2 were investigated by keeping the initial amount of H2O2 constant at 0.2 mM and changing the HCO3– concentration (1 mM, 2 mM, 5mM, and 10 mM). Besides, the effect of UO22+ on the speciation was investigated by adding uranyl nitrate (UO2(NO3)2 x 6H2O) to the systems before exposure to H2O2. The impact of speciation on the kinetics of oxidative dissolution of UO2 was analyzed. As a result of experiments, it has been concluded that the amount of dissolved UO22+ is higher in higher HCO3– concentration. Also, the rate of the UO22+ dissolution decreases with addition of UO22+ due to the complexes formed in the systems. It was observed that oxidation of UO2 is the rate limiting reaction atthe beginning of the exposure; therefore, there is a delay in the UO22+ dissolution. On the other hand, it has been seen that the HCO3– deficiency limits the dissolution capacity of the systems. Free H2O2 is the dominant peroxide species in the systems without initially added UO22+ , while -6 and -2 charged complexes are dominant in the systems with initially added UO22+. The H2O2 complexes are found more effective on the surface mechanism in the systems having lower HCO3– concentration. There is no observable trend in H2O2 consumption rate with respect to HCO3– concentration. Therefore, it was concluded that the H2O2 consumption rate is independent of dissolution reaction. Finally, the dissolution in the system without initially added UO22+ follows the first-order kinetics with respect to HCO3– concentration.

No description available.