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Computational Analysis of Nuclear Fuel of Pressurized Water Reactors
Hřebačka, Michal ; Brně, Ondřej Šťastný, ONZHN SÚJB a FEKT VUT v (referee) ; Katovský, Karel (advisor)
This thesis focuses on calculating the multiplication factor of selected nuclear fuels for pressurized water reactors using two different computational tools. The first tool is a relatively simple code named UWB1 developed at the University of West Bohemia, and the second is the official and widely used computational system SCALE. The results of both programs are compared using a set of fuels used at the Dukovany Nuclear Power Plant with further discussion on their operational and economic characteristics. The final part is dedicated to the analysis of the geometry of fuels offered by final vendors in connection with the Dukovany II project.
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Innovative nuclear fuels and simulation of their burn-up
Pibil, Vojtěch ; Lovecký,, Martin (referee) ; Katovský, Karel (advisor)
The thesis deals with advanced nuclear fuels with increased accident tolerance. The work summarizes the solutions of nuclear fuel manufacturers, describes their problems and advantages over the currently used fuel. In the next part, calculations are performed on uranium tellurides and germanides fuels. The results of these calculations are then used to determine whether any of the tellurium or germanium isotopes behave as a burnable absorber.
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Computational simulation of the third unit of the Dukovany Nuclear Power Plant
Musil, Michal ; Mičian, Peter (referee) ; Katovský, Karel (advisor)
Bachelor‘s thesis deals with the analysis of the possibilities of increasing the output of the 3rd production unit at the Dukovany Nuclear Power Plant. First, the primary circuit of the Dukovany Nuclear Power Plant and the development of various types of nuclear fuel are described. The theoretical part also contains a basic description of deterministic and stochastic calculation methods and a description of the MOBY-DICK calculation program. The practical part contains a comparison of several cases for which an increase of the power output of Unit 3 of the EDU would be possible. All cases are compared in terms of nuclear fuel consumption and the potential for achieving increased power output.
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Accident Tolerant Fuel simulation loaded in advanced nuclear reactor during severe accident conditions
Hamřík, Lukáš ; Foral, Štěpán (referee) ; Mičian, Peter (advisor)
Tato diplomová práce se zaměřuje na paliva se zvýšenou tolerancí vůči haváriím, jež jsou považována za další krok ke zvyšování bezpečnosti provozu jaderných elektráren. Zpočátku jsou odhaleny slabiny v současnosti provozovaného palivového systému. Jsou popsány možnosti pokročilých paliv a jejich kategorizace, v návaznosti na jejich stav vývoje a problémy, kterým čelí před svým nasazením. Je zmíněna elektrárna APR1400 se svými systémy pro zmírnění následků havárie. Model APR1400 je využit k posouzení vlivu pokročilých materiálů pokrytí při těžké havárii. Stručně jsou představeny možnosti a schopnosti kódu MELCOR, který je schopen simulovat danou problematiku těžkých havárií. Je popsána implementace pokročilých materiálů pokrytí do programu MELCOR, společně se simulovanými scénáři. Na závěr jsou diskutovány výsledky získané ze simulací.
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Spent fuel pool accident simulation
Strieš, Jakub ; Foral, Štěpán (referee) ; Mičian, Peter (advisor)
The aim of the thesis is to provide a theoretical basis for simulating spent fuel pool accidents and to try to build a working model in the Melcor environment. The thesis is divided into four main parts. In the first part a description of the spent fuel pool is given. The processes that take place in nuclear fuel and thus lead to the formation of spent fuel are described. The reasons for the need to cool nuclear fuel after it has been in the reactor are described. For example, the generation of residual heat is described. This is followed by a description of the basic parts and purpose of a spent fuel pool. Structural parts, measuring systems or requirements for the pool itself are shown. Subsequently, an overview of accidents in spent fuel pools that have occurred in the past is also included in this chapter. Their causes and consequences are described, but especially their impact on the further development of nuclear safety, especially in the field of spent fuel pools. For both description and subsequent simulation, the two most studied ones, namely the loss-of-cooling failure and the loss-of-coolant failure, have been selected. The third section introduces the Melcor program which, is the main tool to create the simulation model. The chapter contains a basic description of the program as well as an introduction to the modules that are specifically designed for spent fuel pool simulations. The last part is a description of the model used, its creation, modification, problems in its creation and their solution. Finally, the results obtained are shown and described, together with a commentary to complete their understanding.
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Analysis of the innovated nuclear fuel with FEMAXI code
Čásar, Ondřej ; Novotný, Filip (referee) ; Foral, Štěpán (advisor)
This bachelor thesis deals with modification of source code of program FEMAXI 6 and subsequent analysis of composite nuclear fuel based on SiC admixture. First part introduces the entire open fuel cycle, from the start of the mining to the final disposal of the nuclear waste. The following part describes physical processes in the fuel, especially the processes associated with the low thermal conductivity of classical uranium oxide ceramic pellets, contains the description of the temperature calculations in the pellet and the analysis of the thermal conductivity coefficient including the influences affecting its size. One part of the thesis is an analysis of selected composite fuels increasing the thermal conductivity and introduction of a computational program for FEMAXI 6 nuclear fuel analysis, including its structure, implementation of equations and description of the input file. In the practical part, one can find a description of performed program modifications, comparison with other computational programs and analysis of possible composite nuclear fuel for Dukovany NPP.
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Modern nuclear reactors
Šimík, Michal ; Suk, Ladislav (referee) ; Maar, Tomáš (advisor)
This bachelor’s thesis creates an overview of particular types of nuclear reactors, their history, present and future. Large part focus on the objectives required by modern nuclear reactors provided an international forum GIF. Reactors of fourth generation are characterized also their advantages and disadvantages. In the last section is detail description of selected type of reactor.
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Assessment of the thermomechanical behaviour of perspective nuclear fuel for reactivity insertion accidents
Halabuk, Dávid ; Suk, Ladislav (referee) ; Martinec, Jiří (advisor)
The objective of this master’s thesis is to simulate thermo-mechanical behaviour of nuclear fuel in a pressurized water reactor during a reactivity initiated accident. An important part of this work is focused on examination of processes which occur during such accident and on creation of a detailed overview of material properties of nuclear fuel and fuel cladding which are necessary for simulations that closely reflect reality. Simulations in this thesis examine cases of fresh or irradiated nuclear fuel for two types of fuel cladding, Zircaloy-4, a material that is currently used in nuclear reactors, and ceramic matrix composite material made of SiC. The thesis also presents comparison of results with a corresponding international benchmark and an assessment of the influence of selected input parameters on obtained results.
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Enrichment and reprocessing of nuclear fuel
Janoušek, Petr ; Elbl, Patrik (referee) ; Sitek, Tomáš (advisor)
Nuclear fuel is one form of electricity generation. This bachelor thesis deals with its modification and processing as this fuel has to undergo special processes to ensure its good usability. The aim of the thesis is to list and analyze the main processes involved in the production of nuclear fuel. The thesis is, in particular, focused on the enrichment of nuclear fuel, which is the process of enhancing its properties. The first and second chapter of the paper discusses basic issues of first nuclear energy and then nuclear fuel. The third chapter of the thesis explains in detail the enrichment of nuclear fuel and also describes different methods of enrichment. Another process taking place after nuclear fuel is used is it's reprocessing. This method of converting used fuel into fuel that can be reused is dealt with in the last chapter of the thesis.
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Simulation of Accelerator Driven Nuclear Reactor for Spent Nuclear Fuel Transmutation
Jarchovský, Petr ; Ing. Antonín Krása, Ph.D., SCK.CEN Mol (referee) ; Katovský, Karel (advisor)
This master’s thesis deals with usage of burn-up (spent) nuclear fuel in nuclear power plants of next generation – accelerator driven transmutation plants which is produced in current nuclear power plants. This system could significantly reduce the volume of dangerous long-lived radioisotopes and moreover they would be able to take advantage of its great energy potential due to fast neutron spectrum. In the introduction are listed basic knowledge and aspects of spent nuclear fuel along with its reprocessing and the possibility of further use while minimizing environmental impact. As another point detailed description of accelerator driven systems is described together with its basic components. In addition this search is followed by individual chronological enumeration of projects of global significance concerning their current development. Emphasis is placed on SAD and MYRRHA projects, which are used like base for calculations. This last, computational part, deals with the creation of the geometry of subcritical transmutation reactor driven by accelerator and subsequent evaluation which assembly is the most effective for transmutation and energy purposes along with changing of target, nuclear fuel and coolant/moderator.
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