Thermal-hydraulics validation of CFD code for light water nuclear reactors against benchmark experimental results

Authors

  • Deiglys Borges Monteiro Federal University of ABC https://orcid.org/0000-0003-2071-0010
  • Duvan Alejandro Castellanos Gonzalez Federal University of ABC
  • José Rubens Maiorino Federal University of ABC

DOI:

https://doi.org/10.15392/bjrs.v9i2B.1403

Keywords:

Benchmark, CFD, subchannel, fuel rod, validation

Abstract

The cooling of a nuclear reactor depends on a suitable fluid flow pattern among its fuel elements aiming the removal of heat produced in the fuel. In case of light water reactors, an excess of heat drives the fluid to change its phase from liquid to vapor, significantly reducing its capacity to remove heat and leading the reactor to a Loss of Coolant Accident. Numerical simulations using a CFD code is a suitable tool to address this kind of problem and explore the conditions that should be avoided during the reactor operation. The commercial CFD codes had proven to be reliable to simulate with a high accuracy and confidence the thermal-hydraulics of a sort of equipment and systems, avoiding spending efforts and financial resources in the development of new codes that, essentially, perform the same tasks. Despite of it, the CFD codes must be validated, such as against experimental results. To comply with this objective, a benchmark fuel element was purposed and experimentally essayed to provide experimental results for CFD codes calibration. The results of this essay are provided to the four types of subchannels for a 5x5 PWR fuel element, with results provided as density and void fraction. This work presentes the preliminary results obtained with CFD numerical simulations using the ANSYS-CFX® code for the central subchannel with active rods for stead state operation. The results demonstrated that the ANSYS-CFX® is adequate to simulate with high accuracy the flow in this subchannel.

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Author Biographies

Deiglys Borges Monteiro, Federal University of ABC

Programa de pós-graduação em Energia (PPGENE)/ Universidade Federal do ABC/Centro de Engenharia e Ciências Sociais Aplicadas (CECS)/Pró-reitoria de Pesquisa (PROPES)

Duvan Alejandro Castellanos Gonzalez, Federal University of ABC

Programa de pós-graduação em Energia (PPGENE)/ Universidade Federal do ABC/Centro de Engenharia e Ciências Sociais Aplicadas (CECS)

José Rubens Maiorino, Federal University of ABC

Programa de pós-graduação em Energia (PPGENE)/ Universidade Federal do ABC/Centro de Engenharia e Ciências Sociais Aplicadas (CECS)

References

D’AURIA, F. Priorization of nuclear thermal-hydraulics researches. Nuclear Engineering and Design, v.340, p.105-111, 2018.

MESQUITA, A.M.; RODRIGUES, R. R. Detection of the Departure from Nucleate Boiling in Nuclear Fuel Rods Simulators. International Journal of Nuclear Energy, v.2013, Article ID 950129, 7 pages, 2013.

LUTSANYCH, S.; MORETTI, F.; D’AURIA, F.; Validation of NEPTUNE CFD two-phase flow models against OECD/NRC PSBT subchannel experiments. Nuclear Engineering and Design, v.321, p.82-91, 2017.

NEA – NUCLEAR ENERGY AGENCY; OECD/NRC benchmark based on NUPEC PWR sub-channel and Bundle test (PSBT), Volume 1: Experimental Database and Final Problem Specifications. Report. OECD/NEA, Vol. 1, 2012.

PEÑA, C.; PELLACANI, F.; CHIVA, S.; BARRACHINA, T.; MIRÓ, R.; JUAN, R.M.; CFD-Neutronic coupled calculation of a quarter of a simplified PWR fuel assembly including spacer pressure drop and turbulence enhancement. In: Proceedings of INAC 2011, ABEN, Belo Horizonte, Brasil, Oct. 24-28, 2011.

NEA – NUCLEAR ENERGY AGENCY; International benchmark on Pressurized Water Reactor Sub-channel and Bundle Tests, Volume 2: Benchmark results of phase I – Void distribution. Report. OECD/NEA, Vol. 2, 2016.

ANSYS Free Student Software Downloads. ANSYS. Available: https://www.ansys.com/academic/free-student-products. Last accessed: 2019.

KREPPER, E.; RZEHAK, R.; CFD Analysis of a Void Distribution Benchmark of NUPEC PSBT Tests: Model Calibration and Influence of Turbulence Modeling. Science and Technology of Nuclear Installations, v. 2012, Article ID 939561, 10 pages, 2012.

IN, W.K.; SHIN, C.H.; LEE, C.Y.; CFD simulation of subcooled boiling flow in nuclear fuel bundle. In: Proceedings of Seventh International Conference on Computational Fluid Dynamics (ICCFD7), Big Island, Hawaii, USA, July 9-13, 2012.

ANSYS User Guide Manual. CFX. Release 19.2. Canonsburg, PA, 2018.

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Published

2021-07-25

How to Cite

Monteiro, D. B., Gonzalez, D. A. C., & Maiorino, J. R. (2021). Thermal-hydraulics validation of CFD code for light water nuclear reactors against benchmark experimental results. Brazilian Journal of Radiation Sciences, 9(2B (Suppl.). https://doi.org/10.15392/bjrs.v9i2B.1403

Issue

Section

XXI Meeting on Nuclear Reactor Physics and Thermal Hydraulics (XXI ENFIR) and VI ENIN