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Ionizing radiation shielding simulation using MCNP code
Konček, Róbert ; Košťál,, Michal (referee) ; Katovský, Karel (advisor)
Radiation is defined as ionizing if it has enough energy to remove electrons from atoms or molecules when it passes through or collides with matter. This ability implies potentially detrimental effects on living tissue. Ionizing radiation shielding is therefore a discipline of great practical importance. The thesis builds upon the author's previous work on the topic and widens the scope of discussion with theoretical and practical issues of advanced shielding calculations. The theoretical part of the thesis describes several approaches to calculating fluence or absorbed dose at an arbitrary point in space. Point-kernel methods provide sufficiently accurate results for simpler shielding problems. In many practical cases, however, calculations based on the transport theory are necessary. There are two basic types of transport calculations: deterministic transport calculations in which the linear Boltzmann equation is solved numerically, and Monte Carlo calculations in which a simulation is made of how particles migrate stochastically through the problem geometry. Advantages and disadvantages of both methods are discussed. In the practical part are the results of radiation shielding calculations performed with a major Monte Carlo code - MCNP6, compared with those obtained in the experiments, which were carried out at the Ionizing Radiation Laboratory at Department of Electrical Power Engeneering, FEEC BUT. The experiments consisted of placing a cobalt-60 radioisotope source at three different positions inside a lead collimator, and counting pulses with two different scintillation detectors positioned in front of the opening of the collimator, alternately with or without lead shield located between the source and the used detector. Agreement of the calculations and the data from the measurements is reasonable, given the inherent uncertainties of the experimental set-up. Performed sensitivity analysis shows relative importances of different parameters used as inputs in simulations, such as densities of materials, or dimensions of the scintillation crystals. Annotated MCNP input files used for simulation are also part 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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Nuclear reactor power detectors based on Cherenkov radiation measurement
Čejka, Lukáš ; ČEZ,, Ondřej Zlámal, (referee) ; Katovský, Karel (advisor)
This work deals with the connection between the luminance of Cherenkov radiation in a nuclear reactor and his power. Cherenkov radiation arises when charged particle passing through the medium if their speed is higher than the phase velocity of light in that medium. In water moderated reactors, it is caused by secondary electrons produced by absorption of gamma rays. The measurement is done using the camera and software for photometric luminance measurements. Furthermore, there is mentioned the use of Cerenkov radiation detection to measure burnout of spent nuclear fuel. There are examples of usage of Cherenkov radiation detection to measure of nuclear reactors power, including an analysis of the measured results on research nuclear reactor VR -1. The measured results show increasing intensity of Cherenkov radiation with increasing reactor power, but residual detection response caused by delayed gamma radiation was observed after the rapid decrease of the reactor power. Possible use of detection luminance of Cherenkov radiation is in monitoring unforeseen increase in reactor power or in the control of used fuel.
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Possibilities of thorium utilization in current NPPs
Svoboda, Josef ; Praze, Jan Prehradný, ÚE FS ČVUT v (referee) ; Katovský, Karel (advisor)
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.
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Analysis of the Most Common Causes of Nuclear Fuel Failures During Operation
Ježek, Martin ; ČEZ,, Ondřej Zlámal, (referee) ; Katovský, Karel (advisor)
Nuclear fuel failures during the reactor operation happen quite often in the world. The theoretical part of this thesis is dedicated to the most common causes of nuclear fuel failures. It describes failure mechanism and corrective actions. The unfavorable trends in nuclear fuel behavior are prevented by suitable method of nuclear fuel monitoring. Some of them may affect the safety of the power plant. For example, the fuel assembly bow affects the function of rod cluster control assembly. Another part, which describes inspection methods, is devoted to inspection and repair of nuclear fuel. The thesis concentrates on the Temelin NPP, where there was implemented post-irradiation inspection program for checking compatibility between Westinghouse's fuel assemblies and water chemistry of reactor VVER. This program continues even after the change of nuclear fuel supplier. Practical part of this thesis is dedicated to proposal of a new method of fuel assembly bow measurement for Temelin NPP based on ultrasound. This proposal is supported by measurement on the experimental device for detection of spacer grid position developed by Research Centre Rez.
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Computational analysis of pressurized water reactor core behaviour using PARCS code
Novotný, Filip ; Ing. Jan Frýbort, Ph.D., KJR FJFI ČVUT v Praze (referee) ; Katovský, Karel (advisor)
The Master Thesis performs search concerning advanced small and medium power light-water reactors’ designs, including different possibilities to gain a license for their development and operation. The work covers the principal theory about the area of neutronics calculations, principal equations and simplifications. There are several different methods for solution of neutronics calculations. The thesis gives an overview of two principal groups of codes – deterministic methods and Monte Carlo method. The survey shows computational codes examples based on mentioned methods. The computational code PARCS is chosen for further study, which contained description of the input and output file, process of the model creation and conditions for neutronics calculation the of selected reactor design. Based on these facts, the transient calculation has been prepared within the thesis. Thee analyses are described – reactor emergency shutdown, reactor shutdown with stuck group of control and emergency shutdown rods and reactor shutdown with faulty reaction of emergency shutdown rods.
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Evaluation of Nuclear Fuel Safety and Reliability Using Research Reactors' In-Core Experiments
Matocha, Vítězslav ; Foral, Štěpán (referee) ; Katovský, Karel (advisor)
The aim of this master thesis is to show a connection among nuclear fuel safety, experiments led in research reactors and calculation codes. This thesis focuses on the calculation code Transuranus. There are represented four experiments, which were calculated in Transuranus. The fission gas release, elongation and growth of fuel were particularly monitored. Is is possible to set differences among versions v1m1j09 and v1m3j12 from achieved results, as well as the influence of selected Transuranus parameters on the results, so the thesis may bring new pieces of knowledge for improvement of safety analysis calculation by Transuranus.
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Nanotechnology utilization in nuclear industry and research
Skalička, Jiří ; Štefánik, Milan (referee) ; Katovský, Karel (advisor)
This thesis introduces reader to current knowledge of nanomaterials and their usage. It summarises production methods and usage of different materials in nuclear power plants, nuclear research and nuclear medicine. Theoretical part of this thesis is dedicated to possible usage of carbon nanotubes for neutron beam collimation and guides. In experimental part different materials were tested in measuring box connected to horizontal radial channel of VR-1 nuclear reactor and their influence on neutron flux was measured. Tested samples were non-oriented carbon nanotubes, carbon nanofibers, alumina nanowires, oriented carbon nanotubes with several angles of rotation and these samples were compared with results of graphite.
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Project of nuclear power unit secure emergency power supply for design basis accident as well as for extension conditions
Žák, Tomáš ; Šťastný, Jiří (referee) ; Katovský, Karel (advisor)
The present Master´s thesis aims at designing an evolutionary scheme of secured power supply at nuclear power plants with VVER 440 reactors during design basis accidents as well as design extension conditions. In the first part of this thesis, concepts relating to the defence in depth of nuclear reactors, operating modes of the blocks as well as types and possibilities of electrical power supply and electric circuits of the block are defined. Although the present thesis deals with PWR 440 in general, special emphasis is put on the Czech NPP in Dukovany, where there are four PWR 440 reactors in operation, and on the possibilities for enhancing the defence in depth in this area. The second part of the thesis deals with the difference in station blackout definitions before and after Fukushima; not only the differences in situation evaluation are dealt with, but a solution is also proposed to make the system of secured power supply system during design basis accidents as well as design extension conditions more robust. This option has been selected out of a number of possibilities based on the evaluation of reliability, availability and cost-effectiveness of the proposal.
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