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ISSN Online: 2377-424X

ISBN Print: 0-89116-299-2

International Heat Transfer Conference 7
September, 6-10, 1982, Munich, Germany

COMPILATION OF CRITICAL HEAT FLUX DATA IN PWR FUEL ASSEMBLIES WITH NON-UNIFORM AXIAL HEAT FLUX DISTRIBUTION

Get access (open in a dialog) DOI: 10.1615/IHTC7.3500
pages 447-452

Abstract

Critical Heat Flux tests using water as the coolant have been carried out at the Heat Transfer Research Facility of Columbia University over the last 20 years on many types and designs of nuclear reactor fuel assemblies. Herein the geometric characteristics of 52 rod bundle test sections with non-uniform axial heat flux distribution are presented. A total of 2 029 CHF data points from these test sections are stored on magnetic tape for ease of reference and analysis.
The geometries and ranges of operating parameters are typical of pressurized water reactors. Distribution of the data with respect to the geometric and operating parameters of the test sections and the errors associated with the measurements are given.