Fluxo crítico de calor como parâmetro de projeto em sistemas nucleares
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Tipo
TCC
Data de publicação
2019-12-13
Periódico
Citações (Scopus)
Autores
Rezende, Levi Camilo Marques de
Bortotti, Luís Vitor Rodrigues
Bortotti, Luís Vitor Rodrigues
Orientador
Pessanha, Jorge Alexandre Onoda
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Resumo
As usinas nucleares são responsáveis por cerca de 10% da energia elétrica gerada no mundo em 2018, sendo o seu elemento essencial o reator nuclear. O reator central de estudo deste trabalho é o reator de água pressurizada, um dos muitos tipos existentes e o mais utilizado no mundo. O principal parâmetro de projeto desse tipo de reator é o fluxo crítico de calor, conceito que, por intermédio de métodos de modelagem experimental e analítica, estima o máximo de calor que pode ser trocado entre as varetas combustíveis e o líquido refrigerante de modo seguro. Com este, visa-se a aprofundar a explicação do fenômeno do fluxo crítico de calor a partir da investigação de artigos científicos e relatórios oficiais de domínio público, comparando alguns dos métodos existentes e analisando principalmente a correlação empírica W-3. Para a melhor compreensão do tema abordado neste trabalho, é dada uma visão geral do que é energia nuclear, como é usada para a geração de energia, o reator de água pressurizada e suas características, uma explanação aprofundada do que é o fluxo crítico de calor e os métodos usados para o estimar. Foi possível, através deste, explicar o que é fluxo crítico de calor e os mecanismos que o geram, bem como apresentar e comparar os métodos de previsão deste. Foi, também, possível depreender que, por não haver uma equação teórica que modele perfeitamente a crise de ebulição, foram desenvolvidas diversas correlações e métodos empíricos.
Nuclear power plants are responsible for around 10% of all electric energy generated worldwide nowadays, with the nuclear reactor being its core element. The central reactor of this work is the pressurized water reactor, one of the many existing types and the most used worldwide. The main project parameter of this reactor type is the critical heat flux, concept that, by means of experimental and analytical modeling methods, estimates the maximum heat that can be exchanged between the fuel rods and the coolant in a safe way. This aims to deepen the explanation of the critical heat flux based on the investigation of scientific publications and official reports of public domain, comparing some of the existing equations and analyzing mainly the empirical correlation W-3. For the better comprehension of the discussed subject in this work, it is given a general view of what nuclear energy is, how it’s used in generating electricity, the pressurized water reactor and its characteristics, a deepened explanation of critical heat flux and the methods used to estimate it. It was possible, through this, to explain what is critical heat flux and the mechanisms that generate it, as well as presenting and comparing its prediction methods. It was also possible to infer that, in the absence of a theoretical equation that perfectly models the boiling crisis, there were developed many empirical correlations and methods.
Nuclear power plants are responsible for around 10% of all electric energy generated worldwide nowadays, with the nuclear reactor being its core element. The central reactor of this work is the pressurized water reactor, one of the many existing types and the most used worldwide. The main project parameter of this reactor type is the critical heat flux, concept that, by means of experimental and analytical modeling methods, estimates the maximum heat that can be exchanged between the fuel rods and the coolant in a safe way. This aims to deepen the explanation of the critical heat flux based on the investigation of scientific publications and official reports of public domain, comparing some of the existing equations and analyzing mainly the empirical correlation W-3. For the better comprehension of the discussed subject in this work, it is given a general view of what nuclear energy is, how it’s used in generating electricity, the pressurized water reactor and its characteristics, a deepened explanation of critical heat flux and the methods used to estimate it. It was possible, through this, to explain what is critical heat flux and the mechanisms that generate it, as well as presenting and comparing its prediction methods. It was also possible to infer that, in the absence of a theoretical equation that perfectly models the boiling crisis, there were developed many empirical correlations and methods.
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Palavras-chave
fluxo crítico de calor , reator nuclear , modelagem , W-3 , critical heat flux , nuclear reactor , modeling