Thermal-hydraulic analyzes of a slow loss of flow accident in the IEA-R1 Nuclear Reactor using RELAP and CFD codes
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Brazilian Journal of Radiation Sciences
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Among the most critical accidents for the IEA-R1, there is the Loss of Flow Accident (LOFA) in which a sudden and abrupt stop of the primary pump causes the loss of flow. Traditionally, this kind of accident is analyzed by thermal-hydraulic system codes. However, they can overestimate the fluid and fuel temperature along the transient by up to 20%. Moreover, thermal-hydraulic system codes can face difficulties to capture three-dimension phenomenon, such as the natural convection. Meanwhile, Computational Fluid Dynamics (CFD) analysis has shown good results when analyzing open-pool research reactor accidents even dealing with flow inversion. This work presents a thermal-hydraulic analysis of a Slow Loss of Flow Accident (SLOFA) in the IEA-R1 using the RELAP5 code and the commercial CFD code Ansys CFX®. The objective is to combine the advantages of both approaches. The system code was used to find the transient boundary conditions for a CFD model. The CFD software solved the detailed flow pattern in a quarter of a fuel channel. The numerical results showed good agreement with the benchmark data. The peak temperatures were overestimated in only 1.8 °C in the fluid and 3 °C in the cladding.
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COSTA, F.C.; BELCHIOR JUNIOR, A.; TORRES, W.M.; UMBEHAUN, P.E.; ANDRADE, D.A.D. Thermal-hydraulic analyzes of a slow loss of flow accident in the IEA-R1 Nuclear Reactor using RELAP and CFD codes. Brazilian Journal of Radiation Sciences, v. 13, n. 4, p. 1-26, 2025. DOI: 10.15392/2319-0612.2025.2987. Disponível em: https://repositorio.ipen.br/handle/123456789/49676. Acesso em: 29 Apr 2026.
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