Please use this identifier to cite or link to this item: http://www.repositorio.cdtn.br:8080/jspui/handle/123456789/1022
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dc.coverage.spatialBelo Horizonte
dc.date.accessioned2016-08-29T19:21:15Z-
dc.date.available200705
dc.date.available2016-08-29T19:21:15Z-
dc.date.issued2006
dc.identifier.urihttp://www.repositorio.cdtn.br:8080/jspui/handle/123456789/1022-
dc.description.abstractExperimental and analytical studies have been performed in the IPR-R1 TRIGA Mark-1 reactor at Nuclear Technology Development Center (CDTN), Brazil, to find out the temperature distribution under steady-state conditions as a function of the reactor power. The flow distribution in the cooling channels and the heat-transfer coefficient on the heated surface were obtained experimentally. These results permitted to make the prediction of critical heat flux (CHF), which defines the limit of fuel heat removal.
dc.language.isoInglês
dc.publisherCDTN
dc.rightsL
dc.subjectTRIGA- Brazil reactor
dc.subjectcritical heat flux
dc.typeTrabalho Apresentado em Evento
dc.coverage.spatialeventoBelo Horizonte^BMG^CBrazil
dc.creator.affiliationCentro de Desenvolvimento da Tecnologia Nuclear/CDTN, Belo Horizonte, MG, Brasil
dc.creator.affiliationCentro de Desenvolvimento da Tecnologia Nuclear/CDTN, Belo Horizonte, MG, Brasil
dc.date.evento22-25 Aug 2006
dc.identifier.eventoExperimental prediction of the critical heat flux on the IPR-R1 TRIGA nuclear reactor
dc.title.eventoWorld TRIGA users conference, 3
Appears in Collections:Trabalho apresentado em evento

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