Behavior of thorium plutonium fuel on light water reactors

• Autor(a) principal: GOMES, DANIEL S.
• Data de Publicação: 2020
• Outros Autores: SILVA, ANTONIO T. e, OLIVEIRA, FABIO B.V. de, LARANJO, GIOVANNI S., INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE
• Tipo de documento: Artigo de conferência
• Título da fonte: Repositório Institucional do IPEN
• Texto Completo: http://repositorio.ipen.br/handle/123456789/30707
• topic: closed fuel cycle; computerized simulation; delayed neutron fraction; f codes; monte carlo method; nuclear fuel conversion; nuclear fuels; plutonium; reactivity coefficients; thermal conductivity; thorium; uranium dioxide; water cooled reactors

Designs using thorium-based fuel are preferred when used in compliance with sustainable energy programs, which should preserve uranium deposits and avoid the buildup of transuranic waste products. This study evaluates a method of converting uranium dioxide (UO2) to thorium-based fuel, with a focus on Th-Pu mixed oxide (Th-MOX). Applications of Th-MOX for light water reactors are possible due to inherent benefits over commercial fuels in terms of neutronic properties. The fuel proposed, (Th-Pu)O2, can be helpful because it would consume a significant fraction of existing plutonium. Aside from the reactor core, the proposed fuel could be useful in existing technology, such as in a pressurized water reactor (PWR). However, licensing codes cannot support Th-MOX fuel without implementing adaptations capable of simulating fuel behavior using the FRAPCON code. The (Th-Pu)O2 fuel should show a plutonium content that produces the same total energy release per fuel rod when using UO2 fuel. Thorium is a fertile material and demands a slightly higher plutonium content when used in Th-MOX. Mixed ceramic oxides show thermodynamic responses that depend on the comprising chemical fractions, and there is little information in databases on irradiation effects. The neutronic analysis is carried out using the SERPENT code to quantify transuranic production and compare this production with the original UO2 fuel assembly. Parameters such as delayed neutron fraction and temperature reactivity coefficient are also determined. Through these analytical methods, the viability and sustainability of the proposed new fuel assembly can be demonstrated in a closed fuel cycle.