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Research and Development of Generation IV. reactors with an Emphasis on Molten Salt Reactors
Slančík, Tomáš ; Novotný, Filip (referee) ; Katovský, Karel (advisor)
The bachelor thesis reviews the particular generations of nuclear reactors from the first prototypes to nowadays advanced reactors. It also provides an overview of projects supporting development of IV. generation of nuclear reactors. The main part describes the selected reactors of generation IV by the Generation IV International Forum, focusing on molten salt reactors. The last chapter is devoted to a practical part of bachelor thesis focused on simulated calculation of neutron flux in various salts using MCNP program.
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Safety problems of modern nuclear reactors
Kukla, Jaroslav ; Matějíček, Luboš (referee) ; Přibil, Rudolf (advisor)
Jaroslav Kukla Otázky bezpe nosti moderních jaderných reaktor Bakalá ská práce Abstract: This study is concerned with safety factors of nuclear reactors. It summarizes history of nuclear reactors development and analyses causes of some serious accidents, which occurred in the past. It describes physical characteristics of radioactive radiation and mentions a basic survey of quantities and units. Then it describes the influence of radioactive emission on man and other living organisms. It also defines some reactor components and focuses on general security principles, which are further developed. It shortly explains the methods of evaluation of safety factors of nuclear reactors and a role of nuclear supervision. It puts nuclear reactors into several generations and every type of reactor is described as for technical and safety aspects, respectively. It refers to contemporary trends in development of nuclear reactors and in development of their safety.
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History of serious accidents at nuclear facilities
DRAŽANOVÁ, Kristýna
The bachelor thesis maps all the accidents assessed by the INES scale level 5 and above. It is not only concerning events at nuclear power plants, but also at nuclear weapons production facilities, whose construction preceded the development of the nuclear power plants in several countries. An attention is paid also to the unfortunate event, which was originated as a result of unauthorized manipulation with a part of the radiotherapy device containing radioactive isotopes of caesium 137Cs. The theoretical part of the bachelor thesis explains the general concepts, such as radioactivity and ionizing radiation. It distinguishes the types and sources of ionizing radiations and describes its impact on human body. Next, it deals with the concept of nuclear safety, determined supervising authorities and its governing legislation. It also contains a description of the INES scale, which is an international scale used to assess nuclear incidents and accidents. The practical part of this thesis provides an overview of the accidents with a detailed description of their development, the circumstances that led to them and the consequences left behind. The development of the individual accidents is subsequently analysed with a view to discover the underlying causes, and finally compared in order to find possible identical features or similarities. The present thesis was intended to examine the following research question: Is human failure the most frequent cause of nuclear plant accidents?
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Nuclear Fuel Margins Increasing at Nuclear Power Plant Dukovany
Vrábel, Radek ; ČEZ,, Ondřej Zlámal, (referee) ; Katovský, Karel (advisor)
This bachelor‘s thesis deals with power uprates and the extension of the nuclear fuel cycle at Czech nuclear power plants. The beginning of this thesis aims at improving the reader’s knowledge about nuclear reactors and nuclear fuel cycle. A whole chapter is dedicated to the development of nuclear fuel cycle at the nuclear power plant Dukovany. The progress in the field of nuclear fuels complements the development of the nuclear fuel cycle, thus providing visible benefits.
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Advanced nuclear reactors’ promising fuels
Kadlec, Miroslav ; Varmuža, Jan (referee) ; Katovský, Karel (advisor)
This bachelor thesis focuses on the development of nuclear reactors and the fuel burned them. Thesis describes the various types of nuclear fuel, including the ways they are treated either before or after use in a nuclear power plant. Also included is a description of the situation of nuclear power plants in the Czech Republic and the Slovak Republic. In conclusion focuses on the state of some upcoming projects of the future.
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Determination of the temperature reactivity coefficient for Nuclear power plant Dukovany
Klíma, Jakub ; Bajgl, Josef (referee) ; Raček, Jiří (advisor)
This thesis deals with the re-determination of the temperature reactivity coefficient of the Nuclear power plant Dukovany, which is measured within the physical test Reheating of the cooling to 260 °C. This paper contains description of the types of nuclear reactors launching, list of tests which are performed during these startings and it also describes methods for determination of temperature reactivity coefficient. Last part of this thesis includes description of programme which was created for determining temperature reactivity coefficient from given files and verification of its calculating methods by statistical test.
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Accidents of light water reactors
Bejček, Patrik ; Novotný, Filip (referee) ; Foral, Štěpán (advisor)
This thesis deals with the description of primary and secondary circuit of nuclear power plant with reactor VVER-400. Moreover, this thesis describes nuclear safety, its particular factors, and also international nuclear event scale. It is followed by a description of companies playing a significant role in the field of nuclear safety and a close attention is paid to SÚJB. In a section dedicated to nuclear safety, the focus is on events which can occur in nuclear power plant. In addition, the accident Large Break LOCA is described in detail. The practical part deals with the description and instructions to simulator PCTran developed by Micro-Simulation Technology. In the last part, there is an analysis of selected parameters which are simulated in simulator PCTran, parameters simulated in simulator RELAP5 5 and also the calculation of the temperature parameters of fuel TVSA-T and their simulation.
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Benchmarks and test facilities of VVER reactors
Šimek, Ondřej ; Foral, Štěpán (referee) ; Vojáčková, Jitka (advisor)
The aim of this bachelor’s thesis is to describe the topic of benchmarks and test facilities for pressurized water reactors of the eastern concept VVER type. The theoretical part introduces the VVER type reactors, their history, the development of separate generations, including fundamental differences and parameters of the VVER type reactors being used so far. Next two chapters of the theoretical part focus on nuclear safety and security, describing the uppermost Czech authorities, which are SÚJB, MAAE and NEA. Furthermore, many terms concerning nuclear safety are explained in these two chapters. The next chapter is focused on deterministic safety analyzes, their classification, methods and purposes. The aim of this part is also to explain the verification and validation of computing codes. The next chapter offers insight into test facilities which are crucial for evaluation and testing of nuclear devices and computing codes. The last chapter of the theoretical part focuses on the VVER type reactor benchmarks. The practical part of this thesis presents the conversion of a single AER Benchmark FCM_001 using neutron physical code PARCS. The results are compared to the results of CRONOS computing code.
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Depletion calculation of VVER 1000 reactor fuel using KENO code
Janošek, Radek ; Katovský, Karel (referee) ; Novotný, Filip (advisor)
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.
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Provisions for mitigation of consequences in case of major accidents in GFR nuclear reactors
Mlčúch, Adam ; Suk, Ladislav (referee) ; Martinec, Jiří (advisor)
This thesis deals with the severe accident of the gas-cooled fast reactor GFR. At the beginning of the study there is a review of the gas-cooled fast reactor subject. Next part is focused on description of possible solutions for severe accidents with emphasis on the solution applied in the Generation III+ reactors. Chapters that deal with material and thermal balance with severe accident of GFR demonstration unit, along with the chapter which analyses features of the corium, create a basis for the conceptual design of core catcher of GFR demonstration unit, which forms the final part of this thesis.
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