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The Swedish Nuclear Fuel and Waste Management Company (SKB) has developed a method for final disposal of spent nuclear fuel. This technique requires accurate measurement of the residual decay heat of every assembly. For this purpose, depletion codes as well as calorimetric and gamma-ray spectroscopy experimental methods have been developed and evaluated. In this work a prototype analysis tool has been developed to automate the analysis of both calorimetric and gamma-ray spectroscopy measurements. The performance of the analysis tool has been investigated by comparing its output with earlier results and calculations. Parallel to the software development, new measurements on 73 BWR assemblies were performed. The results obtained for the determination of the residual decay heat are presented. Finally, suggestions for further development are outlined and discussed.

In a nuclear fuel cask, the heat generating spent fuel rods are packed in a housing and the resulting bundle is placed inside a cask of thick outer shell made of materials like lead or concrete. The cask presents a wide variation in geometrical dimensions ranging from the diameter of the rods to the diameter of the cask. To make the problem tractable, first the heat generating rod bundle alone is considered for analysis and the effective thermal conductance of the bundle is correlated in terms of the relevant parameters. In the second part, the bundle is represented as a solid of equivalent thermal conductance and the attention is focused on the modelling of the cask. The first part, dealing with the effective thermal conductance is solved using Fluent software, considering coupled conduction, natural convection and surface radiation in the heat generating rod bundle encased in a hexagonal sheath. Helium, argon, air and nitrogen are considered as working media inside the bundle. A correlation is obtained for the critical Rayleigh number which signifies the onset of natural convection. A correlation is also developed for the effective thermal conductance of the bundle, considering all the modes of transport, in terms of the maximum temperature in the rod bundle, pitch-to-diameter ratio, bundle dimension (or number of rods), heat generation rate and the sheath temperature. The correlation covers pitch-to-diameter ratios in the range 1.1-2, number of rods ranging from 19 to 217 and the heat generation rates encountered in practical applications. The second part deals with the heat transfer modeling of the cask with the bundle represented as a solid of effective (or equivalent) thermal conductance. The mathematical model describes two-dimensional conjugate natural convection and its interaction with surface radiation in the cask. Both Boussinesq and non-Boussinesq formulations have been considered for convection. Numerical solutions are obtained on a staggered mesh with a pressure correction method using a custom-made Fortran code. The surface radiation is coupled to the conduction and convection at the solid-fluid interfaces. Steady-state results are obtained using time-marching. Results for various quantities of interest, namely, the flow and temperature distributions, Nusselt numbers, and interface temperatures, are presented. The Grashof number based on the volumetric heat generation and gap width is varied from 105 to 5 ×109. The emissivities of the interfaces are varied from 0.2-0.8 for the radiative calculations. The solid-to-fluid thermal conductivity ratio for the inner cylinder is varied in the range 5-20 in the parametric studies. Simulations are also performed with thermal conductivity calculated in an iterative manner from bundle parameters. The dimensionless outer wall conductivity ratio is chosen to correspond to cask walls made of lead or concrete. The dimensionless thickness (with respect to gap width) of the outer shell is in the range of 0.0825-1, while the inner cylinder dimensionless radius is 0.2. Air is the working medium in the cask for which the Prandtl number is 0.71. Correlations are obtained for the average temperatures and Nusselt numbers at the inner interface in terms of the parameters. The radiation heat transfer is found to contribute significantly to the heat dissipation.

This thesis is focused to the economic analysis of the nuclear fuel strategy change at nuclear power plant Temelin, where the change to 18 months fuel cycle is considered. The introduction of theses is aimed to the identification of direct economic aspects for the financial analysis. Nuclear fuel strategy change affects operation of power plant as a whole and affects production of spent nuclear fuel. Therefore the economic analysis is needed for include all social costs of the change. The conclusion of thesis is aimed to the risk analysis of the nuclear fuel strategy change. Risk analysis is performed by Monte Carlo simulation.

This diploma thesis deals with storing the spent nuclear fuel and reviewing its safety. The theoretical part analyzes the processes taking place while the fuel is being used, such as fission, isotopic changes, fission gas release, cracking, swelling and densification of fuel pellet. The thesis is also focused on handling the spent fuel and on the way it makes from the reactor, through the spent fuel pool, the transportation, various kinds of storing, till the reprocessing and final deep geological repository. Furthermore, this part of the thesis briefly discusses computing code MCNP, its main characteristics, input files and using. The practical part of the work is focused on creating the model of the spent fuel pool located next to the nuclear reactor WWER 440/V213. This type was chosen, because it is the most used type of nuclear reactor in Czech Republic and Slovakia. With the help of the code MCNP, the multiplication factor of the main configurations of the fuel in the pool was calculated, and then the required safety regulations to ensure sufficient subcriticality, so its safety, were checked. Next, several analysis were performed using this model. These analyses were concerning the temperature of coolant, fuel and the use of various nuclear data libraries. In the future this model can be used to realize new analyses with new kinds of fuels, materials and data libraries.

Author´s name: Bc. Stanislav Klimek School: Brno University of Technology, Faculty of Mechanical Engineering, Energy institute Title: Determining the life storage of a carbon steel cask Consultant: Prof. Ing. Oldřich Matal, CSc. Number of pages: 70 Year: 2009 The assignment of this diploma thesis is to estimate the lifetime of spent fuel container made from carbon steel grade. This container is designed for deep geological disposal of spent nuclear fuel. Basic mechanism of corrosion are described in detail in the first part. Further on, this work deals with the other specific phenomena and influences, which affect at corrosion of steel in conditions of a deep geological repository. Heat, radiation and surroundings are considered of particular importance. In the following part an estimate of the lifetime of model container is introduced, which is affected by temperature and radiation. Here recommendations for protection of container are introduced, arising from the model calculation. Finally, the relevancy of incidence of particular parameters is evaluated, which affect the corrosion.

This diploma thesis deals with nuclear energetics and its impact to the environment. Present state of the energetics, mainly nuclear energetics, in the Czech Republic is discussed here. Also perspective of nuclear energetics is given. The thesis describes nuclear power plant waste handling and also spent nuclear fuel handling. Nuclear power plant is compared with other sources of electric energy, which are counted to the energetic mix. The author focuses on fossil fuels and also on the renewable energy sources. The amount of produced waste to the unit of produced energy, built up area, safety and economic indicators and other parameters are followed for each source of energy.

The aim of this work was to study the separate effect of fission fragment damage on the structural integrity and matrix dissolution of uranium dioxide in water. Radiation damage similar to fission damage was created by irradiating bulk undoped and doped ‘SIMFUEL’ disks of UO2, undoped bulk CeO2 and thin films of UO2 and CeO2 with high energy Xe and U ions. The UO2 thin films, with thicknesses in the range of 90 – 150 nm, were deposited onto (001), (110) and (111) orientations of single crystal LSAT (Al10La3O51Sr14Ta7) and YSZ (Yttria-Stabilised Zirconia) substrates. The CeO2 thin films were deposited onto single crystal silicon (001) substrates. Part of the bulk UO2 and CeO2 samples, the thin films of UO2 on the LSAT substrates and the thin films of CeO2 were irradiated with 92 MeV 129Xe23+ ions to a fluence of 4.8 × 1015 ions/cm2 to simulate the damage produced by fission fragments in uranium dioxide nuclear fuel. Part of the bulk UO2 and CeO2 samples and the thin films of UO2 on the YSZ substrates were irradiated with 110 MeV 238U31+ ions to a fluence of 5 × 1010, 5 × 1011 and 5 × 1012 ions/cm2 to study the accumulation of the damage induced. The irradiated and unirradiated samples were studied using scanning electron microscopy (SEM), focused ion beam (FIB), atomic force microscopy (AFM), energy dispersive X-ray (EDX) spectroscopy, electron probe microanalysis (EPMA), X-ray diffraction (XRD), electron backscatter diffraction (EBSD), secondary ion mass spectrometry (SIMS) and X-ray photoelectron spectroscopy (XPS) techniques to characterise the as-produced samples and assess the effects of the ion irradiations. Dissolution experiments were conducted to assess the effect of the Xe ion irradiation on the dissolution of the thin film UO2 samples on the LSAT substrates and the bulk and thin film CeO2 samples. The solutions obtained from the leaching of the irradiated and unirradiated samples were analysed using inductively coupled plasma mass spectrometry (ICP-MS). XRD studies of the bulk UO2 samples showed that the ion irradiations resulted in an increased lattice parameter, microstrain and decreased crystallite size, as expected. The irradiated UO2 thin films on the LSAT substrates underwent significant microstructural and crystallographic rearrangements. It was shown that by irradiating thin films of UO2 with high energy, high fluence ions, it is possible to produce a structure that is similar to a thin slice through the high burn-up structure. It is expected that the ion irradiation induced chemical mixing of the UO2 films with the substrate elements (La, Sr, Al, Ta). As a result, a material similar to a doped SIMFUEL with induced radiation damage was produced.

La caractérisation juste et reproductible en isotopie et en concentration des radioéléments est l’une des thématiques essentielles des laboratoires d’analyse dans le domaine du nucléaire. Afin de minimiser les temps de manipulation en boite à gants des personnels et la production de déchets radioactifs liés à l’analyse de combustibles nucléaires, il est nécessaire de proposer des solutions efficaces et innovantes. Depuis quelques années, la miniaturisation des systèmes séparatifs constitue l’un des axes de développement majeurs de la chimie analytique et ces microsystèmes constituent certainement une des solutions pour répondre aux exigences de l’analyse nucléaire. Ce travail a pour objectif la conception d’une plateforme analytique miniaturisée et à usage unique, dédiée a la séparation des lanthanides, issus des combustibles usés, en amont de leur analyse par spectrométrie de masse. Destiné à remplacer une étape de séparation chromatographique au centre d’un processus analytique de trois étapes, le nouveau protocole basé sur l’isotachophorèse (ITP) doit satisfaire un cahier des charges précis. Les propriétés de complexation des lanthanides ont d’abord été exploitées afin d’obtenir avec un agent chélatant unique et rigoureusement sélectionné, l’acide 2-hydroxy-2-methylbutyrique (HMBA), la sélectivité intra période nécessaire à leur séparation complète par ITP. Basées sur des modèles théoriques existants, des études complémentaires, notamment des paramètres influençant la résolution, ont permis l’amélioration des performances globales du système ainsi que son dimensionnement. Pour réduire drastiquement le volume de déchets liquides secondaires (solutions de rinçage) et la manipulation de matériaux et de matériels radioactifs, le protocole a été implanté sur un microsystème polymérique jetable en COC, spécialement développé pour cette application. Ce microsystème a ensuite été couplé à un spectromètre de masse a multi collection et source à plasma à couplage inductif pour mesurer les rapports isotopiques / The accurate and reproducible characterization of radioactive solutions in isotope composition and concentration is an essential topic for analytical laboratories in the nuclear field. In order to reduce manipulation time in glove box and production of contaminated wastes, it is necessary to propose innovative and efficient solutions for these analyses. Since few years, microchips are a major field of development in analytical chemistry and those devices could provide a solution which fits the needs of nuclear industry. The aim of this work is to design a disposable analytical micro-device devoted to lanthanide separation from spent nuclear fuel before their analysis in mass spectrometry. Designed to be used in place of a separation process by liquid chromatography which is involved in a three step protocol, the new protocol based on isotachophoresis (ITP) keeps compatible with the other two steps. The complete separation of lanthanides by ITP was obtained by the use of only one chelating compound rigorously selected: the 2-hydroxy 2-methyl butyric acid (HMBA). The main parameters involved in solute resolution were defined from the theoretical models of ITP and experimental studies of the influence of these parameters allowed to optimize the geometry of the system and to improve its performances. To suppress cleaning of the system and, consequently, to strongly reduce both liquid waste volume and handling radioactive material, the ITP protocol was transferred in a polymeric (COC) disposable microchip especially developed for this purpose

It is important to know the heat generated due to nuclear decay in the final repository for spent nuclear fuel. In Sweden, the heating powers generated in spent nuclear fuels are currently measured in the calorimeter System 251 at the Clab facility, Oskarshamn. In order to better measure, and increase understanding, of the temperature measurements in the calorimeter, a simple thermodynamic model of its temperature evolution was developed. The model was described as a system of ordinary differential equations, which were solved, and the solution was applied to calibration measurements of the calorimeter. How precise the model is, how its parameters affect the model, et cetera, are addressed. How the temperature evolution of the system changes as the values of parameters in the model are changed is addressed. The mass correction of the calorimeter could be estimated from this model, which validated the established mass correction of the calorimeter. How the measurement results from the calorimeter would be affected if the volume of the calorimeter was changed was also considered. Additionally, gamma radiation escape from the calorimeter without being detected as heat in the calorimeter. The gamma escape energy fraction was estimated by SERPENT simulations of the calorimeter, as a function of the initial photon energy. The gamma escape was also estimated for different values of the radius of System 251.