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Design of the internal circuit of the system for long-term heat removal from NPP hermetic zone
Zemanová, Silvie ; Baláš, Marek (referee) ; Milčák, Pavel (advisor)
The diploma thesis deals with the safety of nuclear power plants and the design of a system for long-term heat removal from the hermetic envelope of the WWER 440 model 213 NPP intended for severe accident management. The thesis contains a description of the course of selected severe accidents and the causes of their occurrence and presents approaches (the concept of defence in depth) and measures (implementation of new systems) by which operators aim to prevent and prepare for their management. The practical part of the diploma thesis is devoted to the design of the piping route of the long-term heat removal system (dimensions, material, wall thickness, layout) and the calculation of the hydraulic losses of this route, it also deals with the conceptual design of the heat exchanger for the above-mentioned emergency system and its thermal calculation.
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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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Nuclear power plants and their safety
Novotný, Petr ; Köbölová, Klaudia (referee) ; Milčák, Pavel (advisor)
The bachelor thesis deals with the current state of nuclear energetics. It describes the development of nuclear power plants to date and specifically describes the most commonly used types of nuclear reactors. Most of the reactors currently in operation are second generation reactors. However, a number of third-generation reactors with improved technological and safety parameters are already under construction. Improved safety lies in the improvement of active safety systems, but also in the increased use of passive safety systems. The paper concludes with a comparison of specific solutions from the most important nuclear reactor manufacturers in the world.
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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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Approach to the nuclaer safety of the 3rd generation nuclear reactors
Pavlíček, Michal ; Kolář, Jaroslav (referee) ; Kolář, Jaroslav (referee) ; Matal, Oldřich (advisor)
The main target of the master´s thesis is reviewing the generation III nuclear reactors in term of the nuclear safety. At first we have to learn some theory of the nuclear safety in order to understand safety systems of the generation III nuclear reactors. Therefore the thesis is divided into two parts. Legislative and technical approaches to nuclear safety are mentioned in the first part. Regulatory bodies, whose task is to supervise nuclear safety in the nuclear power plants, belongs to the legislative approaches. There are defined terms such as defence in depth, redundancy, diversity, etc. There are mentioned methods to assessing nuclear safety – deterministic and probabilistic methods, especially probabilistic methods, for which a simple example is provided. There are also mentioned active and passive safety systems and their significance for nuclear safety and inherent safety too. There is an example of the function of the active and passive safety systems of the EDU nuclear power plant in conclusion of this issue. The second part deals with description of the selected nuclear reactors in context of the construction of the new units of nuclear power plant in Temelín. The nuclear reactors from companies, which applied for the public tender opened by ČEZ, a. s., for the construction of the ETE 3+4. Thus, the nuclear reactor MIR-1200 by ATOMSTROYEXPORT (Russian Federation), the nuclear reactor AP1000 by WESTINGHOUSE (USA) and the nuclear reactor EPR by AREVA (France) are taken into account . Comparison of the generation II and these generation III+ nuclear reactors necessarily belongs to this master´s thesis. These the generation III+ nuclear reactors are compared with the nuclear reactor VVER 440 (EDU) and in particular with the nuclear reactor VVER 1000, which is operated in the nuclear power plant Temelín. The final chapter contains generally appraisal of the whole problem.
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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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Nuclear power plants
Krakovský, Matej ; Špiláček, Michal (referee) ; Milčák, Pavel (advisor)
The bachelor´s thesis explains different types of nuclear reactors and their fuel and safety characteristics. The thesis is complemented with history and description of nuclear powerplants in Slovakia. There are mentioned some of the experimental projects, whose usage is expected in the future. In the beginning are described fundamentals of nuclear energetics and nuclear fission chain reactions.
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Overview and analysis of follow-up actions based on nuclear power plants stress-test results in the Czech Republic and world-wide
Ludvík, Lubomír ; bezpečnost, Ondřej Šťastný, Státní úřad pro jadernou (referee) ; Katovský, Karel (advisor)
This thesis deals with the issue of the stress tests as a reaction to Fukushima accident. Nuclear safety and control systems are generally described in different modes. Fukushima accident is analyzed and described in details. Special attention is focused on the final results and lessons learnt as the background for stress tests goals. The main part of the thesis are the stress tests at all NPP in the European Union overview. Scope of tests was focused on resistance to extreme environmental conditions, a complete loss of safety systems and actions with solution of the severe accidents, including reactor core melting and subsequent effects. The Czech NPPs are discussed in detail, the others in particular way. Finally, recommendations that will be implemented within a few years are summarized. It will strengthen the management of severe accidents beyond the design basic events.
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