Neutronic Evaluation of MSBR System Using MCNP Code

Clarysson Alberto Mello da Silva; Alana Lima Vieira; Isabella Resende Magalhães; Claubia Pereira

doi:10.15392/bjrs.v9i2B.1261

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.

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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.

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