A thermal hydraulic analisys in PWR reactors with UO2 or (U-Th)O2 fuel rods employing a simplified code
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2017
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INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE
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In order to project a nuclear reactor, the neutronic calculus must be validated, so that its thermal limits and
safety parameters are respected. Considering this issue, this research aims to evaluate the APTh
thermal limits. This PWR is a project develope
composed of Uranium and Thorium oxide mixed (U,Th)O2. For this purpose, a simplified, although
conservative, code was developed in a MATLAB environment named
hydraulics Code-Mixed Oxide Thorium”. This code provides axial and radial temperature distribution, as well
as DNBR distribution over the hottest channel of the reactor core. Moreover, it brings other hydraulic quantities,
such as pressure drop over the fuel rod, considering any fuel proportion of (U,Th)O2.The software uses basic
laws of conservation of mass, momentum and energy, it also calculates the thermal conduction equation,
considering the thermal conductive coefficie
finite elements method was used. Furthermore, the proportion of 36% of UO2 was used to evaluate the
temperature over the fuel rod and DNBR minimum in three burn conditions: beginning, middl
program has proven to be efficient in every condition and the results evidenced that the APTh
an initial analysis, has its thermal limits within the recommended security parameters.
Como referenciar
SANTOS, THIAGO A. dos; MAIORINO, JOSE R.; STEFANNI, GIOVANNI L. de. A thermal hydraulic analisys in PWR reactors with UO2 or (U-Th)O2 fuel rods employing a simplified code. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 22-27, 2017, Belo Horizonte, MG. Proceedings... Rio de Janeiro, RJ: Associação Brasileira de Energia Nuclear, 2017. Disponível em: http://repositorio.ipen.br/handle/123456789/28177. Acesso em: 19 Feb 2025.
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