Investigation of discretization uncertainty in Monte Carlo neutron transport simulations of the Molten Salt Fast Reactor (MSFR)

Authors

  • Tiago Augusto Santiago Vieira Centro de Desenvolvimento da Tecnologia Nuclear (CDTN)
  • Felipe Ribeiro Centro de Desenvolvimento da Tecnologia Nuclear (CDTN)
  • Yasmim Carvalho Centro de Desenvolvimento da Tecnologia Nuclear (CDTN)
  • Vitor Silva Jülich Supercomputing Centre (JSC)
  • Graiciany Barros Centro de Desenvolvimento da Tecnologia Nuclear (CDTN)
  • Andre Santos Centro de Desenvolvimento da Tecnologia Nuclear (CDTN)

DOI:

https://doi.org/10.15392/2319-0612.2023.1317

Keywords:

MSFR, Neutronics, Monte Carlo, Extended GCI, Mesh based simulation

Abstract

In the present work, an assessment of the Neutronic Benchmark of the Molten Salt Fast Reactor (MSFR) was performed using mesh based Monte Carlo Neutron Transport (MCNT) calculations with numerical uncertainty quantification due to discretization in neutronic parameters. Calculations with Constructive Solid Geometry (CSG) models where made as a baseline for the developed mesh based models. The numerical uncertainty given by the mesh utilization is evaluated using an extended version of the Grid Convergence Index (GCI). The fuel salt reprocessing is evaluated regarding a constant reprocessing rate. The fuel salt inventory variation with time for the developed models (CSG and meshed) is presented. The differences caused by the discretization procedure are noticeable, which shows that mesh based MCNT require careful mesh sensitivity evaluation and further validation.

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References

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Published

2023-12-19

How to Cite

Vieira, T. A. S., Ribeiro, F. R. C., Carvalho, Y. M., Silva, V. V. A., Barros, G. de P., & Santos, A. A. C. dos. (2023). Investigation of discretization uncertainty in Monte Carlo neutron transport simulations of the Molten Salt Fast Reactor (MSFR). Brazilian Journal of Radiation Sciences, 11(4), 01–27. https://doi.org/10.15392/2319-0612.2023.1317

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