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Title: Experimental determination of heat transfer coefficients in uranium zirconium hydride fuel rod
Title of periodic: International Journal of Nuclear Energy Science and Technology
Authors: Mesquita, Amir Zacarias
Rezende, Hugo César
Affiliation: Centro de Desenvolvimento da Tecnologia Nuclear/CDTN, Belo Horizonte, MG, Brasil
Centro de Desenvolvimento da Tecnologia Nuclear, Belo Horizonte, MG, Brasil
Issue Date: 2007
Keywords: heat transfer;fuel rods;natural convection;subcooled boiling;nucleate boiling;fuel elements;TRIGA- Brazil reactor
Abstract: The heat generated by nuclear fission is transferred from fuel elements to the cooling system through the fuel- to- cladding gap and the cladding-to-coolant interfaces. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were evaluated experimentally. A correlation for the gap conductance between the fuel and the cladding was also presented. As the reactor core power increases, the heat transfer regime from the fuel cladding to the coolant changes from single-phase natural convection to subcooled nucleate boiling. Results indicated that subcooled boiling occurs at the claddingsurface in the central channels of the reactor core at power levels in excess of 265 kW.
Access: L
Appears in Collections:Artigo de periódico

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