Análise termo-hidráulica e neutrônica de um reator ADS

Detalhes bibliográficos
Ano de defesa: 2020
Autor(a) principal: Welen Nunes de Lima
Orientador(a): Não Informado pela instituição
Banca de defesa: Não Informado pela instituição
Tipo de documento: Dissertação
Tipo de acesso: Acesso aberto
Idioma: por
Instituição de defesa: Universidade Federal de Minas Gerais
Brasil
ENG - DEPARTAMENTO DE ENGENHARIA NUCLEAR
Programa de Pós-Graduação em Ciências e Técnicas Nucleares
UFMG
Programa de Pós-Graduação: Não Informado pela instituição
Departamento: Não Informado pela instituição
País: Não Informado pela instituição
Palavras-chave em Português:
MOX
LBE
Link de acesso: http://hdl.handle.net/1843/36597
https://orcid.org/0000-0001-9455-0318
Resumo: In the current scenario of demand for energy free of CO2 emissions and in the possible reduction of the impacts generated by fossil fuels, the nuclear reactors appear as an alternative to obtain electric energy through nuclear fission reactions. For that, it is necessary to develop research on more efficient nuclear reactor technology that uses fuel more efficiently and safely. Within this context, a hybrid multifunctional research reactor for high-tech applications (MYRRHA) is being developed at the Belgian Nuclear Research Center (SCK-CEN), which consists of a fast reactor designed to operate in both the critical and subcritical modes (coupled to an ADS). In order to contribute to the development of these researches in fast reactors, in this dissertation, the MYRRHA reactor was considered for the study of thermohydraulic and neutronic simulations, in steady state, from the computer codes RELAP5-3D and WIMSD-5B respectively and, from such results, simulate the nucleus in several configurations in the NESTLE neutron code, internal to RELAP5-3D, to obtain results such as relative power distribution and effective neutron multiplication factor. For this, constants were calculated for two groups considering an equivalent cell for the entire core; in addition, separate calculations were made for the fuel, coolant and reflector elements in order to obtain the group constants to perform a neutron simulation in the NESTLE code internal to RELAP-3D. Among the main results of the research are presented the parameters generated for two energy groups in the WIMSD code and the values of the neutron multiplication factor in the WIMSD code and in the NESTLE code. The first results demonstrate that the modeling in the WIMSD-5B code is appropriate for the considered simulations and can be implemented to perform more detailed calculations.