One-step coupled calculations (Serpent-OpenFOAM) for a fuel rod of the IPR-R1 triga reactor

Authors

  • Tiago Augusto Santiago Vieira Nuclear Technology Development Center (CDTN)
  • Rebeca Cabral Gonçalves Nuclear Technology Development Center (CDTN)
  • Izabella Cristina de Paiva Machado Nuclear Technology Development Center (CDTN)
  • Guilherme Augusto Moura Vidal Nuclear Technology Development Center (CDTN)
  • Higor Fabiano Pereira Castro Universidade Federal de Minas Gerais https://orcid.org/0000-0003-3822-0546
  • Nataly Lamounier Ribeiro Universidade Federal de Minas Gerais - UFMG
  • Marcos Pais Barroso Filho Nuclear Technology Development Center (CDTN)
  • Wilker Gustavo Ferreira Santos Nuclear Technology Development Center (CDTN)
  • Daniel de Almeida Magalhães Campolina Nuclear Technology Development Center (CDTN)
  • Graiciany de Paula barros Nuclear Technology Development Center (CDTN)
  • Vitor Vasconcelos Araújo Silva Nuclear Technology Development Center (CDTN)
  • André Augusto Campagnole dos Santos Nuclear Technology Development Center (CDTN)

DOI:

https://doi.org/10.15392/bjrs.v9i2B.1295

Keywords:

Monte Carlo, CFD, Multi-physics, Serpent, OpenFOAM.

Abstract

In this work, a single step of coupled calculations for a fuel rod of IPR-R1 TRIGA was performed. The used me-thodology allowed to simulate the fuel pin behavior in steady-state mode for different power levels. The aim of this paper is to present a practical approach to perform coupled calculations between neutronic (Monte Carlo) and thermal-hydraulic (CFD) codes. For this purpose, is necessary to evaluate the influence of the water thermal-physical properties temperature variations on keff parameter. Besides that, Serpent Nuclear Code was used for the neutronics evaluation, while OpenFOAM was used for thermal-hydraulics. OpenFOAM si- mula-tions were made by using a modified chtMultiRegionFoam solver, developed to read Serpent output correctly. The neutronic code was used without any modifications. The results shows that this coupled calculations were consistent and that leads to encouraging further methodology development and its use for full core simulation. Also, the results shows good agreement with calculations performed using other version of OpenFOAM and Milonga as neutronic code.

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Author Biographies

  • Tiago Augusto Santiago Vieira, Nuclear Technology Development Center (CDTN)
    Thermal-Hydraulic and Neutronics Laboratory - LTHN
  • Rebeca Cabral Gonçalves, Nuclear Technology Development Center (CDTN)
    Thermal-Hydraulic and Neutronics Laboratory - LTHN
  • Izabella Cristina de Paiva Machado, Nuclear Technology Development Center (CDTN)
    Thermal-Hydraulic and Neutronics Laboratory - LTHN
  • Guilherme Augusto Moura Vidal, Nuclear Technology Development Center (CDTN)
    Thermal-Hydraulic and Neutronics Laboratory - LTHN
  • Higor Fabiano Pereira Castro, Universidade Federal de Minas Gerais
    Department of Nuclear Engineering
  • Nataly Lamounier Ribeiro, Universidade Federal de Minas Gerais - UFMG
    Department of Mechanical Engineering
  • Marcos Pais Barroso Filho, Nuclear Technology Development Center (CDTN)
    Thermal-Hydraulic and Neutronics Laboratory - LTHN
  • Wilker Gustavo Ferreira Santos, Nuclear Technology Development Center (CDTN)
    Thermal-Hydraulic and Neutronics Laboratory - LTHN
  • Daniel de Almeida Magalhães Campolina, Nuclear Technology Development Center (CDTN)
    Thermal-Hydraulic and Neutronics Laboratory - LTHN
  • Graiciany de Paula barros, Nuclear Technology Development Center (CDTN)
    Thermal-Hydraulic and Neutronics Laboratory - LTHN
  • Vitor Vasconcelos Araújo Silva, Nuclear Technology Development Center (CDTN)
    Thermal-Hydraulic and Neutronics Laboratory - LTHN
  • André Augusto Campagnole dos Santos, Nuclear Technology Development Center (CDTN)
    Thermal-Hydraulic and Neutronics Laboratory - LTHN

References

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http://www.repositorio.cdtn.br:8080/jspui/bitstream/123456789/1220/1/ Tese%20Vitor%20Vasconcelos.pdf, (2016).

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Published

2021-07-25

Issue

Section

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

How to Cite

One-step coupled calculations (Serpent-OpenFOAM) for a fuel rod of the IPR-R1 triga reactor. Brazilian Journal of Radiation Sciences, Rio de Janeiro, Brazil, v. 9, n. 2B (Suppl.), 2021. DOI: 10.15392/bjrs.v9i2B.1295. Disponível em: https://bjrs.org.br/revista/index.php/REVISTA/article/view/1295.. Acesso em: 21 nov. 2024.

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