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Experimental Studies on Critical Heat Flux in Tight Lattice Rod Bundles

[+] Author Affiliations
Xuemei Lang, Houjun Gong, Lei Zhou, Feng Xie, Ye Liu

NPIC, Chengdu, Sichuan, China

Paper No. ICONE22-30915, pp. V02BT09A032; 5 pages
  • 2014 22nd International Conference on Nuclear Engineering
  • Volume 2B: Thermal Hydraulics
  • Prague, Czech Republic, July 7–11, 2014
  • Conference Sponsors: Nuclear Engineering Division
  • ISBN: 978-0-7918-4591-2
  • Copyright © 2014 by ASME


The tight fuel lattice of pressurized water reactors (PWR) is helped to reduce the volume ratio of water-uranium, to increase the conversion ratio, to decrease the volume of core. It is especially useful for very high burnup and high volume power flux. The design of tight-lattice pressurized water reactors requires the knowledge of critical heat flux (CHF) in tight rod bundles. The tight hexagonal 19-rod bundles is used in this test. There are 4 wires wrapped in outside wall of each rod to support and locate. Experimental investigations on CHF behavior in the two kind bundles of helix angle 3° and 5° were performed. The CHF data points have been obtained in a range of parameters: pressure 8.0–16.6 MPa, mass flux 164.6–3283.0 kg/m2s and bundle exit steam quality −0.315 to 0.747. It is found that the CHF value of helix angle 5° bundle was more higher than that of helix angle 3° bundle in the same T/H condition. The effect of different parameters on CHF in the tight rod bundle is similar to that in the open literature. The CHF correlations of helix angle 5° bundle was obtained based on the test data.

Copyright © 2014 by ASME



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