Análise termo-hidráulica e neutrônica de um reator ADS

• Ano de defesa: 2020
• Autor(a) principal: Welen Nunes de Lima
• 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
• Palavras-chave em Português: Análise neutrônica; Análise térmica; MOX; MYRRHA; LBE; RELAP5-3D; LMFBR; NESTLE; Engenharia nuclear; Análise por ativação nuclear
• Link de acesso: http://hdl.handle.net/1843/36597; https://orcid.org/0000-0001-9455-0318

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.