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Главная Архив Thermal Letter Оргкомитет Будущие конференции AIHTC

ISSN Print: 2377-424X

International Heat Transfer Conference 4
August 31 - September 5, 1970, Paris-Versailles, France

A THERMAL-HYDRAULIC SUBCHANNEL ANALYSIS FOR ROD BUNDLE NUCLEAR FUEL ELEMENTS

Get access DOI: 10.1615/IHTC4.2460
pages 1-11

Аннотация

This paper presents a method for calculating the flow and enthalpy in the subchannels of rod bundle nuclear fuel elements. The method uses a mathematical model that considers the effect of cross flow mixing between the flow subchannels. The model's equations are solved by using finite differences and by supplying the turbulent component of cross flow mixing. The validity of the method is shown by using experimental values of turbulent cross flow as input to calculations that successfully predict the flow and enthalpy obtained in an independent experiment.
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