A thermal hydraulic analisys in PWR reactors with UO2 or (U-Th)O2 fuel rods employing a simplified code

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dc.contributor.authorSANTOS, THIAGO A. dos
dc.contributor.authorMAIORINO, JOSE R.
dc.contributor.authorSTEFANNI, GIOVANNI L. de
dc.coverageInternacionalpt_BR
dc.creator.eventoINTERNATIONAL NUCLEAR ATLANTIC CONFERENCEpt_BR
dc.date.accessioned2018-01-02T11:36:32Z
dc.date.available2018-01-02T11:36:32Z
dc.date.eventoOctober 22-27, 2017pt_BR
dc.description.abstractIn 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.pt_BR
dc.event.siglaINACpt_BR
dc.identifier.citationSANTOS, 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. <b>Proceedings...</b> Rio de Janeiro, RJ: Associação Brasileira de Energia Nuclear, 2017. Disponível em: http://repositorio.ipen.br/handle/123456789/28177.
dc.identifier.urihttp://repositorio.ipen.br/handle/123456789/28177
dc.localRio de Janeiro, RJpt_BR
dc.local.eventoBelo Horizonte, MGpt_BR
dc.publisherAssociação Brasileira de Energia Nuclearpt_BR
dc.rightsopenAccesspt_BR
dc.subjectfinite element method
dc.subjectfuel rods
dc.subjectheat transfer
dc.subjectm codes
dc.subjectmixed oxide fuels
dc.subjectnucleate boiling
dc.subjectpwr type reactors
dc.subjects codes
dc.subjectthermal hydraulics
dc.subjectthorium oxides
dc.subjectthorium oxides
dc.subjectthorium oxides
dc.subjecturanium dioxide
dc.titleA thermal hydraulic analisys in PWR reactors with UO2 or (U-Th)O2 fuel rods employing a simplified codept_BR
dc.typeTexto completo de eventopt_BR
dspace.entity.typePublication
ipen.autorGIOVANNI LARANJO DE STEFANI
ipen.codigoautor7606
ipen.contributor.ipenauthorGIOVANNI LARANJO DE STEFANI
ipen.date.recebimento18-01pt_BR
ipen.event.datapadronizada2017pt_BR
ipen.identifier.ipendoc24002pt_BR
ipen.notas.internasProceedingspt_BR
ipen.type.genreArtigo
relation.isAuthorOfPublicationcfb52090-3a5f-42c9-bf50-e969bca3d779
relation.isAuthorOfPublication.latestForDiscoverycfb52090-3a5f-42c9-bf50-e969bca3d779
sigepi.autor.atividadeSTEFANNI, GIOVANNI L. DE:7606:-1:Npt_BR
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