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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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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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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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Thermal and creep analysis of VVER-1000 reactor pressure vessel at high temperatures caused by fuel melting during severe accident
Gabriel, Dušan ; Gál, P. ; Kotouč, M. ; Dymáček, Petr ; Masák, Jan ; Kopačka, Ján
Thermal and creep analysis of the VVER-1000 reactor pressure vessel (RPV) was performed at high temperatures caused by fuel melting during severe accident. First, the integral code ASTEC was applied simulating severe accident evolution since an initiating event up to a hypothetical radioactive release into the environment. The ASTEC outputs including the remaining RPV wall thickness, the heat flux achieved and the temperature profile in the ablated vessel wall served as boundary conditions for the consequent assessment of RPV integrity carried out with the aid of finite element method (FEM). The FEM analysis was performed including the creep behaviour of RPV material using a complex creep probabilistic exponential model with damage. The objective of the analysis was to computationally assess emergency condition and, on this basis, to propose a general methodology for evaluating the integrity of RPV at high temperatures due to fuel melting during severe accident.
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Severe accidents of third generation pressurized water reactors
Nečas, Karel ; Foral, Štěpán (referee) ; Černý, Tomáš (advisor)
The bachelor thesis focuses on the development of global nuclear energy. It clearly describes the development and division of nuclear facilities used for electricity production. The theoretical part of the thesis describes the division of nuclear facilities according to the development generations, the types are discussed in terms of technology and principle of operation. The operating conditions of nuclear reactors are described in detail, with emphasis on emergency conditions. The next part of the thesis focuses on the LOCA accident type for third generation pressurized water reactors. This thesis also includes an analysis of different approaches od the computation codes used for modelling various conditions of the nuclear power plants. The practical part of the bachelor thesis describes nuclear reactor APR1400. It provides its elementary parameters. The calculation code MELCOR was used for modelling this type of the nuclear reactor. The sensitivity analysis of the severe accident was performed using selected parameters. This analysis describes the influence of these parameters on the hydrogen production, the process of damaging of the active zone, and the reactor vessel melting time.
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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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