Thermal-Hydraulic Analyzes of a Slow Loss of Flow Accident in the IEA-R1 nuclear reactor using RELAP and CFD Codes

Franklin Cândido Costa; Antônio Belchior Júnior; Walmir Maximo Torres; Pedro Esnesto Umbehaun; Delvonei Alves D. Andrade

doi:10.15392/2319-0612.2025.2987

Authors

Franklin Cândido Costa Nuclear Energy Research Institute , Naval Systems Design Center
Antônio Belchior Júnior Nuclear Energy Research Institute
Walmir Maximo Torres Nuclear Energy Research Institute
Pedro Esnesto Umbehaun Nuclear Energy Research Institute
Delvonei Alves D. Andrade Nuclear Energy Research Institute

DOI:

https://doi.org/10.15392/2319-0612.2025.2987

Keywords:

CFD, RELAP5, IEA-R1, Thermal-Hydr

Abstract

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.

Downloads

Download data is not yet available.

References

[1] TEIXEIRA e SILVA, A.; MAPRELIAN, A; RODRIGUES, A. C. I.; CABRAL, E. L. L.; MOLNARY, L. D.; MESQUITA, R. N.; MENDONÇA, A. G. “Análise de Segurança do Reator IEA-R1 a 5 MW”, In: Congresso Geral de Energia Nuclear, 8.; Encontro Nacional de Física de Reatores e Termo-Hidráulica, 12., 15-20 out, 2000, Rio de Janeiro, RJ. Anais, Rio de Janeiro: ABEN, 2020. [Online]. Available in: http://repositorio.ipen.br/handle/123456789/13378.

[2] SCURO, N. L.; ANGELO, G.; ANGELO, E.; PIRO, M. H. A.; UMBEHAN, P. E.; TORRES, W. M.; ANDRADE, D. A. D. “Computational Fluid Dynamics Analysis of an Open-Pool Nuclear Research Reactor Core for Fluid Flow Optimization Using a Channel Box”, Nuclear Science and Engineering, vol. 197, no 6, p. 1100–1116, jun. 2023, https://doi.org/10.1080/00295639.2022.2142437. DOI: https://doi.org/10.1080/00295639.2022.2142437

[3] UMBEHAUM, P. E.; ANDRADE, D. A. D.; TORRES, W. M. Torres; RICCI FILHO, W. “IEA-R1 Nuclear Reactor: Facility Specification and Experimental Results”, International Atomic Energy Agency, Vienna, Austria, 2015.

[4] SIEFKEN J., L.; CORYELL, E. W.; HAVERGO, E. A.; HORORST, J. K. “RELAP5 Code Manual, Volume I: Code Structure, System Models, And Solution Methods”, U.S. Nuclear Regulatory Commission Nureg/Cr-5535, Washington, DC, 1995.

[5] D’AURIA, F. Thermal hydraulics in water-cooled nuclear reactors. Oxford: Woodhead publishing, 2017.

[6] HAINOUN, A.; DOVAL, A.; UMBEHAUN, P.; CHATZIDAKIS, S.; GHAZI, N.; PARK, S.; MLADIN, M.; SHOKR, A. “International benchmark study of advanced thermal hydraulic safety analysis codes against measurements on IEA-R1 research reactor”, Nuclear Engineering Design, vol. 280, p. 233–250, dez. 2014, https://doi.org/10.1016/j.nucengdes.2014.06.041. DOI: https://doi.org/10.1016/j.nucengdes.2014.06.041

[7] CAMPOS, R. C. D.; BELCHIOR JUNIOR, A.; SOARES, H. V.; UMBEHAUN, P. E.; TORRES, W. M.; ANDRADE, D. A. D. “Análise de temperaturas em um elemento combustível do reator de pesquisas IEA-R1 durante evento de perda lenta de vazão com RELAP”, Peer Review, vol. 5, no 18, p. 245–267, ago. 2023. DOI: https://doi.org/10.53660/866.prw2310

[8] WANG, M.; WANG, Y.; TIAN W.; QIU, S.; SU, G. H. “Recent progress of CFD applications in PWR thermal hydraulics study and future directions”, Annals of Nuclear Energy, vol. 150, p. 107836, jan. 2021, https://doi.org/10.1016/j.anucene.2020.107836. DOI: https://doi.org/10.1016/j.anucene.2020.107836

[9] SCURO, N. L. “Simulação numérica de um acidente tipo perda lenta de vazão em um reator nuclear de pesquisa”, Mestrado em Tecnologia Nuclear - Reatores, Universidade de São Paulo, São Paulo, 2019. doi: https://doi.org/10.11606/D.85.2019.tde-25102019-154117. DOI: https://doi.org/10.11606/D.85.2019.tde-25102019-154117

[10] ANSYS CFX® 25.0, “User Manual”, ANSYS Europe Ltd., 2025.

[11] SALAMA, A. “CFD investigation of flow inversion in typical MTR research reactor undergoing thermal–hydraulic transients”, Annals of Nuclear Energy, vol. 38, no 7, p. 1578–1592, jul. 2011, https://doi.org 10.1016/j.anucene.2011.03.005. DOI: https://doi.org/10.1016/j.anucene.2011.03.005

[12] KHATER, H.; EL-MORSHEDY, S. E.-D.; ABDELMAKSOUD, A. “Simulation of unprotected LOFA in MTR reactors using a mix CFD and one-d computation tool”, Annals of Nuclear Energy, vol. 83, p. 376–385, set. 2015, https://doi.org/10.1016/j.anucene.2015.03.041. DOI: https://doi.org/10.1016/j.anucene.2015.03.041

[13] HEDAYAT, A.; DAVARI, A. “Feasibility study to increase the reactor power at natural convection mode in Tehran Research Reactor (TRR) through a hybrid thermal-hydraulic simulation and analysis using the RELAP5 code and Computational Fluid Dynamic (CFD) modeling by ANSYS-FLUENT”, Progress in Nuclear Energy, vol. 150, p. 104285, ago. 2022, https://doi.org/10.1016/j.pnucene.2022.104285. DOI: https://doi.org/10.1016/j.pnucene.2022.104285

[14] AUMILLER, D.L.; TOMLINSON, E.T.; BAUER, R.C. A coupled RELAP5-3D/CFD methodology with a proofof-principle calculation, Nuclear Engineering and Design, Volume 205, Issues 1–2, 2001, Pages 83-90, ISSN 0029-5493, https://doi.org/10.1016/S0029-5493(00)00370-8. DOI: https://doi.org/10.1016/S0029-5493(00)00370-8

[15] YE, L.; YU, H.; WANG, M.; WANG, Q.; TIAN, W.; QIU, S.; SU, G.H. CFD/RELAP5 coupling analysis of the ISP No. 43 boron dilution experiment, Nuclear Engineering and Technology, Volume 54, Issue 1, 2022, Pages 97-109, ISSN 1738-5733, https://doi.org/10.1016/j.net.2021.07.047. DOI: https://doi.org/10.1016/j.net.2021.07.047

[16] HA, T.; GARLAND W. J. “Hydraulic study of turbulent flow in MTR-type nuclear fuel assembly”, Nuclear Engineering Design, vol. 236, no 9, p. 975–984, 2006, doi: https://doi.org/10.1016/j.nucengdes.2005.10.004. DOI: https://doi.org/10.1016/j.nucengdes.2005.10.004

[17] NUCLEAR AND ENERGY RESEARCH INSTITUTE (IPEN), “Relatório de Análise de Segurança IEA-R1: Capítulo 16 - Análise de Segurança”, 2017.

[18] CELIK, I. B.; GHIA, U.; ROACHE, P. J., FREITAS, C. J.; COLEMAN, H.; RAAD. “Procedure for Estimation and Reporting of Uncertainty Due to Discretization in CFD Applications”, Journal of Fluids Engineering, vol. 130, no 7, p. 078001, 2008, https://doi.org/10.1115/1.2960953. DOI: https://doi.org/10.1115/1.2960953

[19] INTERNATIONAL ATOMIC ENERGY AGENCY (IAEA), “Research Reactor Core Conversion Guidebook - Voume 2: Analysis”, Vienna 1992.

[20] NUCLEAR AND ENERGY RESEARCH INSTITUTE (IPEN), “Documento de Qualificação - Elemento Combustível Padrão IEA-208”, Centro Do Combustível Nuclear - CCN, jan. 2010.

[21] WAGNER, W.; PRUß, A. “The IAPWS Formulation 1995 for the Thermodynamic Properties of Ordinary Water Substance for General and Scientific Use”, Journal of Physical and Chemical Reference Data, vol. 31, no 2, p. 387–535, jun. 2002, https://doi.org/10.1063/1.1461829. DOI: https://doi.org/10.1063/1.1461829