Avaliação termo-hidráulica para combustível nuclear de alto desempenho utilizando o código de subcanais STHIRP

• Ano de defesa: 2019
• Autor(a) principal: Wallen Ferreira de Souza
• Tipo de documento: Dissertação
• Tipo de acesso: Acesso aberto
• Idioma: por
• Instituição de defesa: Universidade Federal de Minas Gerais; UFMG
• Palavras-chave em Português: STHIRP; Combustível anelar; Termo-hidráulica; Código de subcanais; Modelo térmico; Combustíveis nucleares Elementos; Engenharia nuclear; Combustíveis para reatores nucleares
• Link de acesso: http://hdl.handle.net/1843/RAOA-BCUJEN

The STHIRP-1 code was developed as an activity of the research line related to the thermal hydraulics area of reactors of the Department of Nuclear Engineering of Universidade Federal de Minas Gerais. Is the result of an effort to develop a code that has the same analytical capacity as those developed in institutions and research centers qualified in the nuclear area worldwide. The analytical capacity of the program was tested with the simulation of the system represented by the TRIGA IPR-R1 research reactor installed in the CDTN / CNEN in Belo Horizonte. In this context, it was decided to evaluate the annular fuel, to test the thermal conduction model implemented in the program. The fuel with annular geometry presented in the reference report published by the MIT Center for Advanced Nuclear Energy Systems in 2006 was simulated. This report discusses the proposal for an annular fuel, cooled internally and externally, in order to increase the power density of a PWR reactor without compromising the safety margins of the installation. The thermo-hydraulic conditions were calculated with the aid of the VIPRE. The calculations were always made to follow the information contained in the reference report. The results showed that, in general, there is a good correlation between those predicted by STHIRP-1 and those presented in the report. In the model of a dipstick the good agreement in the comparison of the heat flow, DNBR and pressure drop of VIPRE and STHIRP confirm the native simulation capacity of the annular fuel with the code STHIRP. From the 1/8 core model, representing the complete core by symmetry with a power of 150% in relation to the use of the solid fuel, good agreement of the DNBR (Departure From Nucleate Boiling), and heat flow was obtained. Besides, the radial temperature distributions were obtained and axial direction of the fuel rod and the water temperature. The load loss was the parameter, which showed the greatest difference, in the work. The internal channel had a lower pressure drop, because it does not have a spacer grid. Finally, this study shows that the STHIRP program is a valuable tool that is available to the Department of Nuclear Engineering at no cost and generates perspectives for future work. Ring fuel is very promising, but more research and investment is still needed.