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Title: CHF prediction in nuclear fuel elements by using roud tube data
Title of periodic: Annals of Nuclear Energy Oxford
Authors: Fortini, Maria Auxiliadora
Veloso, Marcelo Antônio
Affiliation: Universidade Federal de Minas Gerais, Belo Horizonte, MG, Brasil
Centro de Desenvolvimento da Tecnologia Nuclear, Belo Horizonte, MG, Brasil
Issue Date: 2002
Keywords: Critical heat flux;fuel elements;fuel assemblies;PWR type reactors
Abstract: The 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.
Access: L
Appears in Collections:Artigo de periódico

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