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Role of Czech NPP within grid power regulation
Kostečka, Jan ; Martinec, Jiří (referee) ; Katovský, Karel (advisor)
The aim of the Bachelor’s thesis is the involvement of Czech nuclear power plants in the Czech energy networks, but also in the European Network and their adaptation. Effect of expansion of renewable electrical energy produces the asymmetry in the network and places formed with an excess of electrical energy and places with lack of electrical energy. Flows of electric energy can affect the security and safety of distribution lines. Excessive overloading can cause them to collapse. Therefore, this work is focused on solutions and avoids this problem and minimizes this potential damage. However, the impact of energy revolution has a giant proportion about these overflows network energy. Uncontrollable wind and solar power source may create regionally region surge and flow of this energy goes to areas with shortages of electrical power. The regulation has an impact on fuel. This renegade production could extend fuel cycle, or expand the campaign to more years at least a year. The calculations can show the amount of fuel that can still be used, and thereby increase the yield and economic benefit conditions for extending the life of the reactor units.
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Nuclear reactors for new units in the Czech Republic
Brunčiaková, Miriama ; Martinec, Jiří (referee) ; Katovský, Karel (advisor)
This bachelor thesis is concerned with nuclear reactors in operation, in construction and in project plans. Because of that it is a very actual theme. Czech Republic tries to solve the question about building two new reactors for Temelín-3,4. There are two competitors in this tender. First of them is the CzechRussian consortium MIR.1200 with the alike named project MIR.1200, and the second one is Westinghouse company with the project AP 1000. Both of these projects belong to III.+ generation of nuclear reactors. In comparison with the previous generations, this generation brings evolution changes and uses much more passive security elements. The project AP 1000 solves heavy accidents by passive security systems solely. The project MIR.1200 uses combination of both active and passive security systems.
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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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Increasing the flow of cooling water for absorption cooling systems in the nuclear power plant Dukovany
Dvořák, Josef ; Suk, Ladislav (referee) ; Martinec, Jiří (advisor)
The thesis focuses on comparison of the original and the new solutions of cooling water circuit of the York cooling units for the purpose of cooling water flow increase for the absorption units in the Dukovany Nuclear Power Plant. The individual parts of the cooling units that were changed and modified within a reconstruction are described here. The aim of the work is also to process and compare the original and the new solutions of the cooling units and the cold source station from the available measured data. The data are processed into illustrative graphs and tables. Based on the obtained data we can observe the changes achieved by the reconstruction that have affected the effectiveness of the cooling units.
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Modernization the Dukovany Nuclear Power Station
Kissler, Martin ; Martinec, Jiří (referee) ; Fiedler, Jan (advisor)
Thesis focuses on a detailed technical description of all important parts of secondary circuit in Dukovany power plant and its connection to other systems of power plant. In thesis are analyzed significant adjustments which have been made during the entire operation of power plant including in particular the actions associated with project called Utilization of project reserves of units EDU. In the main part of the thesis were carried out calculations of the power plant's power for states before and after the modernization and there is also analyzed the impact of individual changes on the whole power plant. These changes are with the entire secondary circuit drawn in the T-s diagram.
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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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Heat transfer in the pressure cover of the GFR reactor
Koryčanský, Roman ; Suk, Ladislav (referee) ; Martinec, Jiří (advisor)
This paper describes the design of the ventilation pressure cover of the demonstrator GFR. The first part is a brief research project GFR and the effects of temperature on the inside structures. In the following part is calculated balance heat losses within the pressure cover for three cases: non-insulated, fully-insulated and partially insulated surface of the reactor. The following is a design of a heat sink for partially insulated surface.
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Possibilities of the external cooling of a pressure vessel in case of the accident with active zone melting
Hanuš, Jan ; Suk, Ladislav (referee) ; Martinec, Jiří (advisor)
The accident at the Fukushima Daiichi nuclear power plant has shown us that there may be situations where the applied technology will not be able to successfully cool the reactor core. These situations may occur when more elements such as supply of energy to power the pumps and diesel generators are destroyed for example by tsunami or earthquake, or other not expected natural disasters. The inability of the residual heat removal leads to the melting of core, relocation to the bottom of reactor pressure vessel (RPV) and failure of RPV. Result of this accident may be containment failure and leakage of fission products into the environment. One way to prevent this scenario may be a passive system called IVR (In-Vessel Retention) by using external cooling of RPV that retains melted core in. This system counts with flooding of RPV´s shaft by water. After natural circulation of water provides the heat transfers from the wall of RPV. The applicability of IVR for VVER 1000 reactors is still in the course of research. However it´s already clear that the submersion of RPV shaft by water will not sufficient. Other elements as suitable insulation and RPV coating which provides a more intensive heat transfer from the walls of RPV will be needed.
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The design of components for emergency cooling of reactor pressure vessel
Katzer, Milan ; Šen, Hugo (referee) ; Martinec, Jiří (advisor)
My thesis deals with the design of an experimental emergency cooling device of the reactor pressure vessel (RPV). It consists of two parts, the theoretical one and practical one. Different molten corium cooling methods in terms of their efficiency and comparison are introduced in the theoretical part. The design of an experimental emergency cooling device, which incorporates a model channel past the reactor pressure vessel , is presented in the practical part. The cooling device consists of a model channel past the reactor pressure vessel, condensator, which takes away the heat generated by the reactor pressure vessel and the pump of a secondary loop. Next, thermal and hydraulic calculations are given in this section. The conclusion is devoted to the evaluation of particular cooling technologies and their comparison in terms of nuclear and technical safety.
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