|
|
Accident Tolerant Fuel simulation loaded in advanced nuclear reactor during severe accident conditions
Hamřík, Lukáš ; Foral, Štěpán (oponent) ; Mičian, Peter (vedoucí práce)
This diploma thesis introduces accident tolerant fuels that are considered to be a next step in increase of safety margins of nuclear power plants. In the beginning, the weaknesses of the currently operated fuel system are exposed. Accident-tolerant fuel options and their categories are described in their stages of development and faced issues before their deployment. The APR1400 plant with its accident mitigation systems is mentioned, the plant model is used to assess the impact of advanced cladding materials in a plant undergoing severe accident. The MELCOR code, capable of simulations of severe accident phenomena, is briefly introduced with its capabilities. The implementation of advanced cladding materials to MELCOR is described, altogether with simulated scenarios. In the end, the results obtained from the simulations are discussed.
|
|
|
Spent fuel pool accident simulation
Strieš, Jakub ; Foral, Štěpán (oponent) ; Mičian, Peter (vedoucí práce)
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.
|
|
|
Těžké havárie tlakovodních reaktorů třetí generace
Nečas, Karel ; Foral, Štěpán (oponent) ; Černý, Tomáš (vedoucí práce)
Bakalářská práce se zabývá vývojem světové jaderné energetiky. Popisuje přehledně vývoj i dělení jaderných reaktorů určených k výrobě elektrické energie. Práce teoreticky popisuje rozdělení jaderných zařízení podle vývojových generací. Rozebrány jsou dané typy z pohledu technologie a principu fungování. Podrobněji jsou popsány provozní stavy jaderných reaktorů s důrazem na havarijní stavy. Další část práce seznámí s typem havárie LOCA pro tlakovodní reaktory třetí generace. Definován je samotný rozbor problematiky výpočtových kódů určených pro modelování havarijních stavů jaderných zařízení. Praktická část bakalářské práce popisuje jaderný reaktor APR1400 a uvádí jeho základní parametry, podrobněji seznamuje s modelem daného reaktoru vytvořeném pomocí výpočtového kódu MELCOR. Navazující část obsahuje citlivostní analýzu vybraných parametrů v modelu při scénáři těžké havárie, kdy dojde k havárii LOCA společně se selháním všech systémů vysokotlakého doplňování chladiva. Rozebírá jejich vliv na množství vytvořeného vodíku, průběh poškození aktivní zóny a čas protavení dna reaktorové nádoby.
|
|
|
A Nodalization Study of Phebus FPT-1 Experiment in MELCOR 2.2
Cerny, T. ; Ptacek, M.
Phebus, an integral experimental facility located in France, is scaled down version of a French 900 MWe pressurized water reactor (PWR) by a ratio of 1/5000. The goal of the facility is to study phenomena occurring during light water reactor severe accidents and validate models and computer codes used for calculation of sever accidents. The aim of this paper is to compare different modelling approaches in MELCOR 2.2 computer code and the influence on thermal hydraulic behavior and core degradation in comparison with experimental results.
|
host ::
RSS.