SCURO, NIKOLAS L.ANGELO, GABRIELANGELO, E.TORRES, WALMIR M.UMBEHAUN, PEDRO E.ANDRADE, DELVONEI A. de2020-01-152020-01-15SCURO, NIKOLAS L.; ANGELO, GABRIEL; ANGELO, E.; TORRES, WALMIR M.; UMBEHAUN, PEDRO E.; ANDRADE, DELVONEI A. de. Preliminary numerical analysis of the flow distribution in the core of a research reactor. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. <b>Proceedings...</b> Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5667-5674. Disponível em: http://repositorio.ipen.br/handle/123456789/30732.http://repositorio.ipen.br/handle/123456789/30732The thermal-hydraulic safety analysis of research reactors establishes the safety criteria to ensure the integrity of the fuel elements in the reactor core. It assures that all core components are being adequately cooled during operation. It is necessary to know if the average mass flow rate (and their standard deviation) among the fuel assemblies are enough to cool the power generated during operation. Once satisfied such condition, it allows the calculation of the maximum heat flux transferred from fuel assemblies to the coolant, and if the maximum cladding temperatures are below the limits set by the safety criteria. Among the objectives, this study presents a methodology for a preliminary three-dimensional numerical analysis of the flow distribution in the core of the IEA-R1 research reactor, under steady state condition. For this, the ANSYS-CFX® commercial code was used to analyze the flow dynamics in the core, and to visualize the velocity field. It was possible to conclude that a homogeneous flow distribution for all standard fuel assemblies were found, with 2.7% deviation from the average mass flow. What turned out to be negligible and can be assumed that there is a homogeneous distribution in the core. Complex structures were find in the computational domain. Once known the core flow dynamics, it allows future studies to determine whether the heat flux and temperature conditions abbeys thermal-hydraulic safety criteria.5667-5674openAccessboundary conditionsc codesflow modelsfuel assembliesiear-1 reactornumerical analysisreactor coresresearch reactorssafetysteady-state conditionsthermal hydraulicsTexto completo de evento0000-0002-6689-3011https://orcid.org/0000-0002-6689-3011https://orcid.org/0000-0002-2887-0759