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.

Gegenstand des Vorhabens im Rahmen der WTZ mit Russland ist die Versprödung des Reaktordruckbehälters infolge der Strahlenbelastung mit schnellen Neutronen im kernnahen Bereich. Um den Einfluss von bestrahlungsinduzierten Gitterdefekten auf die mechanischen Eigenschaften zu ermitteln, wurden analytische Berechnungen zum Einfluss von Hindernissen auf die Beweglichkeit von Versetzungen und damit auf die Ausbildung einer plastischen Zone an der Rissspitze durchgeführt. Es wird demonstriert, dass sich die an der Rissspitze entstehenden Versetzungen an dem Hindernis (bestrahlungsinduzierte Punktdefekte) aufstauen. In Abhängigkeit der Rissbelastung KI und der Entfernung des Hindernisses von der Rissspitze werden die Versetzungsdichte und das durch den Versetzungsstau verursachte Spannungsfeld berechnet. Mit Hilfe von Experimenten zur Neutronenkleinwinkelstreuung (SANS - small angle neutron scattering) an verschiedenen WWER-Stählen und Modelllegierungen wurden Größenverteilungen und die Volumenanteile der strahleninduzierten Defekte für verschiedene Bestrahlungszustände (Fluenzen, Bestrahlungstemperaturen) ermittelt. Es wurde gezeigt, dass sich die strahleninduzierte Werkstoffschädigung durch Wärmebehandlung weitgehend wieder ausheilen lässt. Nach der thermischen Ausheilung ist der Werkstoff bei erneuter Bestrahlung weniger anfällig für strahleninduzierte Defekte. Die Ergebnisse der SANS-Untersuchungen wurden mit der Änderung der mechanischen Eigenschaften (Härte, Streckgrenze und Sprödbruchübergangstemperatur) korreliert. Mit der kinetischen Gitter-Monte-Carlo-Methode wurden numerische Sensitivitätsstudien zum Einfluss des Cu-Gehalts auf die Stabilität von Defekt-Clustern durchgeführt. Die Berechnungen zeigen, dass die Anwesenheit von Cu-Atomen zur Bildung von langlebigen Defekten führt. Dabei werden Leerstellen in Cu/Leerstellen-Cluster eingefangen. Leerstellen in reinem Eisen sind bei Bestrahlungstemperaturen von 270 °C dagegen nicht stabil, die Lebensdauer liegt zwischen 0.01 s und 1 s. Die kritische Cu-Konzentration, ab welcher stabile Defekte entstehen, beträgt ca. 0.1 Masseprozent.

The main aim of this thesis is the model design and control the secondary circuit block VVER 440. The search part of my work is a description of the secondary circuit. In the first part of the calculation is performed for the rated secondary circuit. The second calculation part focuses on the calculation circuit of the secondary for the reduced, influenced by shutting down one steam generator.

The aim of this thesis is to create a conceptual design of the secondary and tertiary circuit of a new nuclear power plant in our country, which is preceded by a brief search of particular generations of nuclear reactors. The VVER-1200 reactor became the model for the new nuclear power plant. There was designed a heat diagram containing primary components in this work. After that, there was the balance calculation and calculation of parameters of the main devices of the proposed scheme such as condenser, low-pressure and high-pressure regeneration exchangers and steam generator.

This Master’s thesis focuses on calculation of power and temperature distribution in fuel assemblies of VVER-440 reactor at Dukovany Nuclear Power Plant. Theoretical section contains a brief description of VVER-440 technology, fuel and its development at Dukovany Nuclear Power Plant, basics of heat generation in nuclear reactors as well as an overview and categorization of computer codes, used for core calculations. Of these codes, the MOBY-DICK computer code is then described in depth, including its input and output files. The MOBY-DICK code is later on used for pinwise calculating power distribution of selected fuel cycles of defined units at Dukovany Nuclear Power Plant, with vizualization of output values for characteristic fuel assemblies. Results of this computation are then used for analysis, whether uneven power distribution in the core and heat generation gradient within fuel assemblies have any influence on measuring channel output temperatures, which is the pivotal part of this thesis.

This thesis deals with the replacement of the lid gasket of the primary collector of the steam generator VVER 1000. Original sealing by nickel rings is replaced by kammprofile gasket with expanded graphite layers. The thesis compare the properties of both types of gaskets and the new and the original configuration of flange joint which have been calculated according to EN 1591. The results are compared and conclusions arising therefrom are presented. The work includes results of FEM analysis of the new configuration of the flange joint. There is also a description of the main components of the nuclear power plants VVER 1000 primary circuit in the introductory part which focused on the construction and operation of the steam generator and its primary collector.

Subject of this thesis is design of the steam generator for modular reactor. The dissertation consist of the theoretical part, where are described heat-exchangers and steam generators used in nuclear power plants. Second part contains theoretical calculations in the first approach (thermal, hydraulic and strenght calculation). In the next part are particular variants of steam generator selected. For the final variant are necessary calculations (introduced above) and drawings of selected parts are done. In the final statement is technical solution evaluated, and the parameters of the steam generator are compared with actually used steam generators.

Nuclear power plants provide about 11 percent of the world's electricity production. For fission process is uranium fuels used with varying percentage of enrichment 235U for most of nuclear reactors. Uranium reserves are reducing and their mining cost increases. Therefore, the thorium fuel is discussed as revolution fuel for current and future nuclear power plants. This diploma thesis deals with possibility of thorium fuel utilization at various types of nuclear reactors with a focus on light water reactors. The practical part of the thesis is focused on simulation and calculations of various uranium dioxide and thorium dioxide layers at the fuel rods. Model of WWER 440 reactor was developed for the calculations with the addition of thorium fuel. The model simulates burning out of fuel for 5 years, with monitoring of fuel behavior and tracking changes of each material. The thesis tries to define the suitable ratio and parameters of layers combination of uranium and thorium fuel. For these ratios and parameters the thesis tries to give sufficient amount of computational analyzes.

The introduction to operational nuclear reactors focusing on light-water pressurized reactor VVER 1000 is in the beginning of this Master´s thesis. This thesis covers basic technology of VVER 1000 reactor with focus on reactor core and nuclear fuel TVSA-T. A significant part of the thesis deal with basic concepts of nuclear safety and its methods. The main goal is to create a model of VVER 1000 reactor, which can be used in nuclear burn-up calculations using KENO code. Therefore a part of this thesis deals with explanation of statistical Monte Carlo method and the KENO code.