Neutronic Evaluation of MSBR System Using MCNP Code

Authors

  • Clarysson Alberto Mello da Silva Departamento de Engenharia Nuclear Universidade Federal de Minas Gerais
  • Alana Lima Vieira Departamento de Engenharia Nuclear Universidade Federal de Minas Gerais
  • Isabella Resende Magalhães Departamento de Engenharia Nuclear Universidade Federal de Minas Gerais
  • Claubia Pereira Departamento de Engenharia Nuclear Universidade Federal de Minas Gerais

DOI:

https://doi.org/10.15392/bjrs.v9i2B.1261

Keywords:

MSBR, Nuclear Fuel Cycle, Neutronic Simulation, MCNPX

Abstract

The concept of Molten Salt Reactor use Th to breed fissile 233U, where an initial source of fissile material needs to be provided. However, there is no available 233U and so; the fissile fuel supply is one of the unresolved problems. Thus, it is necessary to use existing fissile materials such as 235U or Pu to produce 233U. Current studies analyze the fuel transition from 235U/Th or Pu/Th to 233U/Th and, in this context, the present work evaluates the criticality and the neutron flux of MSBR (Molten Salt Breeder Reactor) considering the fuel: (i) mix of Th and enriched U; (ii) the combination of Th and reprocessed Pu; and (iii) matrix of reprocessed Pu/minor actinides (MAs) and Th. The goal is to verify which of these fuels can be used as initial fissile supply. The MSBR core was simulated by MCNPX 2.6.0 code and the criticality model presents similar behavior of previous studies. The results show that reprocessed fuels could have a potential to be used as initial fissile supply, but these fuels present a neutron flux profile less flattens than traditional 233U/Th. It is possible that a new distribution of fuel elements may improve this profile and future simulations will be performed to evaluate this behavior. The uranium, must has high enrichment value to be used as initial seed.  Other studies need be performed to evaluates the uranium enrichment and the U/Th ratio that produces similar core criticality to traditional fuel.

Downloads

Download data is not yet available.

Author Biography

Clarysson Alberto Mello da Silva, Departamento de Engenharia Nuclear Universidade Federal de Minas Gerais

Professor Adjunto da Universidade Federal de Minas Gerais (UFMG) onde desenvolve atividades de ensino e de pesquisa no Departamento de Engenharia Nuclear (DENU). Possui Doutorado (2009) e Mestrado (2005) em Ciências e Técnicas Nucleares pela UFMG, Especialização em Ensino de Física - Universidade Federal de Ouro Preto (UFOP, 2000) e Graduação em Matemática (Licenciatura Plena) - Faculdades Integradas Newton Paiva (FINP, 1996). Possui experiência na área de Engenharia Nuclear com ênfase em Tecnologia Nuclear, atuando em temas relacionados a simulação de sistemas nucleares, reatores avançados, ciclo do combustível, transmutação nuclear e reprocessamento do combustível queimado.

References

SERP, J.; ALLIBERT, M.; BENES, O.; DELPECH, S.; FEYNBERG, O.; GHETTA, V.; HEUER, D.; HOLCOMB, D.; IGNATIEV, V.; KLOOSTERMAN, J. L.; LUZZI, L.; MERLE-LUCOTTE, E.; UHLÍR, J.; YOSHIOKA, R.; ZHIMIN, D. The molten salt reactor (MSR) in generation IV: Overview and perspectives. Progress in Nuclear Energy, v. 77, p. 308-3019, 2014.

RYKHLEVSKIIA A.; BAEA, J. W.; KATHRYN, D. H. Modeling and simulation of online reprocessing in the thorium-fueled molten salt breeder reactor. Annals of Nuclear Energy, v. 128, p .366-379, 2019.

CUI, D.Y.; LI, X. X.; XIA, S. P.; ZHAO, X. C.; YU, C. G.; CHEN, J. G.; CAI, X. Z. Possible scenarios for the transition to thorium fuel cycle in molten salt reactor by using enriched urani-um. Progress in Nuclear Energy, v. 104, p. 75-84, 2018.

ZOU, C.Y.; CAI, C. Z.; YU, C. G.; WU, J. H.; CHEN, J. G. Transition to thorium fuel cycle for TMSR. Nuclear Engineering and Design, v. 330, p. 420-428, 2018.

WEI, H.; CHEN, Y.; LAN, K.; CHENG, J. Parametric study of thermal molten salt reactor neutronics criticality behavior. Progress in Nuclear Energy, v. 108, p. 409-418, 2018.

ZHUANG, K.; CAO, L. Numerical analysis on the dynamic behaviors of a graphite-moderated molten salt reactor based on MOREL2.0 code. Annals of Nuclear Energy, v. 117, p. 3-11, 2018.

LI, G.C.; Optimization of Th-U fuel breeding based on a single-fluid double-zone thorium mol-ten salt reactor. Progress in Nuclear Energy, v. 108, p. 144-151, 2018.

RYKHLEVSKII, A.; LINDSAY, A.; HUFF, K. Full-Core Analysis of Thorium-Fueled Molten Salt Breeder Reactor Using the SERPENT 2 Monte Carlo Code. In: TRANSACTIONS OF THE AMERICAN NUCLEAR SOCIETY, 2017, Washington, D.C., v. 117, p. 1343-1346, 2017.

BETZLER, B. R.; POWERS, J. J.; WORRALL, A. Molten salt reactor neutronics and fuel cycle modeling and simulation with SCALE. Annals of Nuclear Energy, v. 101, p. 489-503, 2017.

PARK, J.; JEONG, Y,; LEE, H. C.; LEE, D. Whole core analysis of molten salt breeder reactor with online fuel reprocessing. International Journal of Energy Research, v. 39, p. 1673-1680, 2015.

ORNL - Oak Ridge National Laboratory. Conceptual Design Study of a Single-Fluid Monten-Salt Breeder Reactor. ORNL Report 4541, EUA, 1971. 207p.

COTA, S.; PEREIRA, C. Neutronic Evaluation of the Non-Proliferating Reprocessed Nu-clear Fuels in Pressurized Water Reactors. Annals of Nuclear Energy, v. 24, p. 829-834, 1997.

LANL – Los Alamos National Laboratory. MCNPX User’s Manual, Version 2.6.0. LANL Report CP-07-1473, EUA, 2008.

Downloads

Published

2021-07-25

How to Cite

Silva, C. A. M. da, Vieira, A. L., Magalhães, I. R., & Pereira, C. (2021). Neutronic Evaluation of MSBR System Using MCNP Code. Brazilian Journal of Radiation Sciences, 9(2B (Suppl.). https://doi.org/10.15392/bjrs.v9i2B.1261

Issue

Section

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

Similar Articles

You may also start an advanced similarity search for this article.

Most read articles by the same author(s)

<< < 1 2