Critical heat flux in horizontal and vertical 37-element nuclear fuel bundles


Atkinson, John Clayton




The flow models for critical heat flux in flow boiling in tubes and rod bundle geometries are reviewed. Sadeghlou's C6D correlation for vertical 37-element bundles at 1000 psia is reviewed and the significance of the Becker-type quality correction factors for geometry/ pressure/ radial flux depression/ heat flux and bundle junction end plates are examined. A CHF correlation for vertical 37-element bundles is presented which describes the available high mass flux data with an rms error of 4.3%.

The effect of buoyant forces on CHF in horizontal flow tubes is reviewed. A brief review of the principles of similitude and modeling along with a description of their application to Freon-to-water modeling in CHF experiments is given. Ahmad's [7] method of compensated distortion is described and his scaling flow parameter fervertical orientations presented. Meri lo' s [8] development of adimensionless scaling flow parameter for horizontal tubes is described. Using the techniques of Ahmad and Merilo, and CHF data from 37-element bundles cooled by Freon and by water, a scalin9 flow parameter for horizontal conditions in 37-element bundles is developed. Using this scaling flow parameter, the calculation of water-equivalent results from Freon-12 CHF data is described. A CHF correlation for horizontal 37-element CANDU bundles is present~d, which describes the water-equivalent data and the actual water data in the range of interest with a rms error of 4.4%. The application of this CHF correlation to the design of CANDU fuel channels is discussed and the benefits of the use of this correlation over the current approach are stated.


Nuclear Fuel Elements -- Thermal Properties




Carleton University

Thesis Degree Name: 

Master of Engineering: 

Thesis Degree Level: 


Thesis Degree Discipline: 

Engineering, Mechanical

Parent Collection: 

Theses and Dissertations

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