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dc.contributor.authorFortini, Maria Auxiliadora
dc.contributor.authorVeloso, Marcelo Antônio
dc.date.accessioned2016-08-29T18:47:33Z-
dc.date.available2204
dc.date.available2016-08-29T18:47:33Z-
dc.date.issued2002
dc.identifier.issnISSN 0306-4549
dc.identifier.urihttp://www.repositorio.cdtn.br:8080/jspui/handle/123456789/500-
dc.description.abstractThe 1995 CHF table for uniformly heated round tubes, developed jointly by Canadian and Russian researchers, has been used for the prediction of critical heat flux (CHF) in 5x5 test sections simulating fuel elements of pressurized water reactors. Comparisons between measured and calculated CHF indicates that the table with an appropriate diameter correction can be applied to rod bundles of the type considered in this study. The relation for the diameter correction factor was derived from the CHF data. The tolerance limits associated with the departure from nucleate boiling ratio (DNBR) are evaluated by using statistical analysis.
dc.rightsL
dc.subjectCritical heat flux
dc.subjectfuel elements
dc.subjectfuel assemblies
dc.subjectPWR type reactors
dc.titleCHF prediction in nuclear fuel elements by using roud tube data
dc.typeArtigo Periódico
dc.creator.affiliationUniversidade Federal de Minas Gerais, Belo Horizonte, MG, Brasil
dc.creator.affiliationCentro de Desenvolvimento da Tecnologia Nuclear, Belo Horizonte, MG, Brasil
dc.identifier.vol29
dc.identifier.extentp.2071-2085
dc.title.journalAnnals of Nuclear Energy Oxford
Appears in Collections:Artigo de periódico

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