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Science and Technology of Nuclear Installations
Volume 2015 (2015), Article ID 243867, 10 pages
http://dx.doi.org/10.1155/2015/243867
Research Article

CHF Enhancement of Advanced 37-Element Fuel Bundles

Korea Atomic Energy Research Institute, 989-111 Daedukdaero, Yuseong-gu, Taejon 305-353, Republic of Korea

Received 28 October 2014; Accepted 30 January 2015

Academic Editor: Antonio Carlos Marques Alvim

Copyright © 2015 Joo Hwan Park et al. This is an open access article distributed under the Creative Commons Attribution License, which permits unrestricted use, distribution, and reproduction in any medium, provided the original work is properly cited.

Linked References

  1. D. C. Groeneveld and K. C. Goel, “A method of increasing critical heat flux in nuclear fuel bundles,” CRNL-1763, 1978.
  2. A. G. McDonald and S. C. Sutradhar, “CANFLEX bundle thermalhydraulic experiments: part 4, Freon CHF tests on the 37E-hybrid bundle, equipped with two space planes and four bearing pad planes,” HPBP-32/ARD-TD-124, 1988.
  3. D. C. Groeneveld, “On the definition of critical heat flux margin,” Nuclear Engineering and Design, vol. 163, no. 1-2, pp. 245–247, 1996. View at Publisher · View at Google Scholar · View at Scopus
  4. G. C. Dimmick, W. W. R. Inch, J. S. Jun et al., “Full scale water CHF testing of the CANFLEX bundle,” in Proceedings of the 6th International Conference on CANDU Fuel, pp. 103–113, Ontario, Canada, 1999.
  5. L. K. H. Leung, J. S. Jun, G. R. Dimmick, D. E. Bullock, W. W. R. Inch, and H. C. Suk, “Dryout power of a CANFLEX bundle string with raised bearing pads,” in Proceeding of the 7th International Conference on CANDU Fuel, pp. 27–40, Kingston, Ontario, Canada, 2001.
  6. L. K. H. Leung and F. C. Dimayuga, “Measurements of critical heat flux in CANDU 37-element bundle with a steep variation in radial power profile,” Nuclear Engineering and Design, vol. 240, no. 2, pp. 290–298, 2010. View at Publisher · View at Google Scholar · View at Scopus
  7. Fuel Design Manual for CANDU-6 Reactors, DM-XX-37000-001, AECL, 1989.
  8. J. H. Bae and J. H. Park, “The effect of a CANDU fuel bundle geometry variation on thermalhydraulic performance,” Annals of Nuclear Energy, vol. 38, no. 9, pp. 1891–1899, 2011. View at Publisher · View at Google Scholar · View at Scopus
  9. M. B. Carver, J. C. Kiteley, R. Q.-N. Zhou, S. V. Junop, and D. S. Rowe, “Validation of the assert subchannel code: prediction of critical heat flux in standard and nonstandard CANDU bundle geometries,” Nuclear Technology, vol. 112, no. 3, pp. 299–314, 1995. View at Google Scholar · View at Scopus
  10. J. H. Park and Y. M. Song, “The effect of inner ring modification of standard 37-element fuel on CHF enhancement,” Annals of Nuclear Energy, vol. 70, pp. 135–140, 2014. View at Publisher · View at Google Scholar · View at Scopus
  11. A. Tahir, Y. Parlatan, M. Kwee, W. Liauw, G. Hadaller, and R. Fortman, “Modified 37-element bundle dryout,” in Proceedings of the 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH '11), Toronto, Canada, 2011.
  12. S. R. Douglas, “WIMS-AECL release 2-5d users manual,” COG-94-52(Rev. 4), FFC-RRP-299, AECL, 2000. View at Google Scholar
  13. C. L. Wheeler, C. W. Stewart, R. J. Cena et al., “COBRA-IV-I: an interim version of COBRA for thermal-hydraulic analysis of rod-bundle nuclear fuel elements and cores,” Battelle Pacific Northwest Laboratories Report BNWL-1962, Battelle Pacific Northwest Laboratories, 1976. View at Google Scholar
  14. C. W. Stewart, C. L. Wheeler, R. J. Cena, C. A. McMonagle, J. M. Cuta, and D. S. Trent, “COBRA-IV: the model and the method,” Battelle Pacific Northwest Laboratories Report BNWL-2214, 1977. View at Publisher · View at Google Scholar
  15. Y. F. Rao, Z. Cheng, G. M. Waddington, and A. Nava-Dominguez, “ASSERT-PV 3.2: advanced subchannel thermalhydraulics code for CANDU fuel bundles,” Nuclear Engineering and Design, vol. 275, pp. 69–79, 2014. View at Google Scholar
  16. D. C. Groeneveld, L. K. H. Leung, P. L. Kirillov et al., “The 1995 look-up table for critical heat flux in tubes,” Nuclear Engineering and Design, vol. 163, no. 1-2, pp. 1–23, 1996. View at Publisher · View at Google Scholar · View at Scopus