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Science and Technology of Nuclear Installations
Volume 2012, Article ID 465059, 12 pages
Research Article

Validation of the Subchannel Code SUBCHANFLOW Using the NUPEC PWR Tests (PSBT)

Institute for Neutron Physics and Reactor Technology (INR), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Germany

Received 19 March 2012; Accepted 5 June 2012

Academic Editor: David Novog

Copyright © 2012 Uwe Imke and Victor Hugo Sanchez. 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. M. J. Thurgood et al., “COBRA/TRAC—A Thermal-Hydraulic Code for Transient Analysis of Nuclear Reactor Vessel and Primary Coolant Systems,” NUREG/CR-3046, 1983.
  2. C. Y. Payk et al., “Analysis of FLECHT SEASET 163-Rod Blocked Bundle Data using COBRA-TF,” NRC/EPRI/Westinghouse-12, 1985.
  3. M. Avramova, Development of an innovative spacer grid model utilizing computational fluid dynamics within a subchannel analysis tool [Ph.D. thesis], Penn State University, 2007.
  4. K. Ivanov and M. Avramova, “Challenges in coupled thermal-hydraulics and neutronics simulations for LWR safety analysis,” Annals of Nuclear Energy, vol. 34, no. 6, pp. 501–513, 2007. View at Publisher · View at Google Scholar · View at Scopus
  5. D. Basile, R. Chierici, M. Beghi, E. Salina, and E. Brega, “COBRA-EN, an Updated Version of the COBRA-3C/MIT Code for Thermal-Hydraulic Transient Analysis of Light Water Reactor Fuel Assemblies and Cores,” Report 1010/1 (revised 1.9.99), ENEL-CRTN Compartimento di Milano.
  6. H. Blasius, “Das Ähnlichkeitsgesetz bei Reibungsvorgängen in Flüssigkeiten,” Forschungs-Arbeit des Ingenieur-Wesens 131, 1913 (in German).
  7. A. A. Armand, The Resistance During the Movement of a Two-Phase System in Horizontal Pipes, vol. 828 of AERE-Lib/Trans, Izvestiya Vsesojuznogo Teplotekhnicheskogo Instituta, 1946.
  8. P. W. Dittus and L. M. K. Boelter, “Heat transfer in automobile radiators of the tubular type,” University of California Publications in Engineering, vol. 2, no. 13, pp. 443–461, 1930, reprinted in International Communications in Heat and Mass Transfer, vol. 12, pp. 3—22, 1985. View at Google Scholar
  9. C. L. Wheeler et al., “COBRA-IV-I: an Interim Version of COBRA for Thermal-Hydraulic Analysis of Rod Bundles Nuclear Fuel Elements and Cores,” BNWL-1962, Battelle, Pacific Northwest Laboratory, RichIand, Washington, DC, USA, 1976.
  10. R. W. Bowring, “Physical model based on bubble detachment and calculation of steam voidage in the subcooled region of a heated channel,” Tech. Rep. HPK-10, Institutt for Atomenergi, Halden, Norway, 1962. View at Google Scholar
  11. H. Dorra, S. C. Lee, and S. G. Bankoff, “A Critical Review of Predictive Models For The Onset of Significant Void in Forced-Convection Subcooled Boiling,” WSRC-TR-93-404, Westinghouse Savannah River Corporation, 1993.
  12. B. Chexal, G. Lellouche, J. Horowitz, and J. Healzer, “A void fraction correlation for generalized applications,” Progress in Nuclear Energy, vol. 27, no. 4, pp. 255–295, 1992. View at Google Scholar · View at Scopus
  13. S.-H. Ahn and G.-D. Jeun, “Effect of spacer grids on CHF at PWR operating conditions,” Journal of the Korean Nuclear Society, vol. 33, no. 3, pp. 283–297, 2001. View at Google Scholar
  14. S. G. Beus, “A Two-Phase Turbulent Mixing Model for Flow in Rod Bundles,” Bettis Atomic Power Laboratory, WAPD-T-2438, 1970.
  15. M. Glück, “Validation of the sub-channel code F-COBRA-TF. Part II. Recalculation of void measurements,” Nuclear Engineering and Design, vol. 238, no. 9, pp. 2317–2327, 2008. View at Publisher · View at Google Scholar · View at Scopus
  16. A. Rubin, A. Schoedel, M. Avramova, H. Utsuno, S. Bajorek, and A. Velazquez-Lozada, “OECD/NRC Benchmark Based on NUPEC PWR Subchannel and Bundle Tests (PSBT), Volume I: Experimental Database and Final Problem Specifications,” US NRC OECD Nuclear Energy Agency, 2010.
  17. D. S. Rowe, “COBRA IIIC: A Digital Computer Program for Steady-State and Transient Thermal Analysis of Rod Cundle Nuclear Fuel Elements,” BNWL-1695, Pacific Northwest Laboratory, 1973.
  18. C. L. Wheeler et al., “COBRA-IV-I: An Interim Version of COBRA for Thermal Hydraulic Analysis of Rod Bundle Nuclear Fuel Elements and Cores,” BNWL-1962, Pacific Northwest Laboratory, 1976.
  19. U. Imke, V. Sanchez, and R. Gomez, “SUBCHANFLOW: an empirical knowledge based subchannel code,” in Proceedings of the Annual Meeting on Nuclear Technology, pp. 4–6, Berlin, Germany, May 2010.
  20. V. Sánchez, U. Imke, and R. Gomez, “SUBCHANFLOW: a thermal-hydraulic sub-channel program to analyse fuel rod bundles and reactor cores,” in Proceedings of the 17th Pacific Basin Nuclear Conference, pp. 24–30, Cancún, México, October 2010.
  21. L. S. Tong, Boiling Heat Transfer and Two-Phase Flow, John Wiley & Sons, 1965.
  22. A. Berkhan, V. Sánchez, and U. Imke, “Validation of PWR relevant models of SUBCHANFLOW using the NUPEC PSBT Data,” in Proceedings of the 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14), pp. 25–29, Toronto, Canada, September 2011.
  23. J. T. Rogers and R. G. Rosehart, “Mixing by turbulent interchange in fuel bundles. Correlations and interfaces,” in: ASME, 72-HT-53, 1972.
  24. J. T. Rogers and A. E. E. Tahir, “Turbulent interchange mixing in rod bundles and the role of secondary flows,” in ASME, Heat Transfer Conference, San Francisco, Calif, USA, 1975.
  25. D. G. Reddy and C. F. Fighetti, “Parametric Study of CHF Data Volume 2, A generalized subchannel CHF correlation for PWR and BWR fuel assemblies,” EPRI-NP-2609, 1983.
  26. M. Calleja, V. Sanchez, and U. Imke, “Implementation of SUBCHANFLOW in the SALOME Platform and Coupling with the Reactor Dynamic Code COBAYA3,” Jahrestagung Kerntechnik, Berlin, Germany, 17.19. Mai, 2011.
  27. A. Gomez, V. Sanchez, U. Imke, and R. Macian, “DYNSUB: A Best Estimate Coupled System for the Evaluation of Local Safety Parameter,” Jahrestagung Kerntechnik, Berlin, Germany, 17.19. Mai, 2011.
  28. Ivanov, V. Sanchez, and U. Imke, “Application of a coupled code system NCNP-SUBCHANFLOW to study conventional and innovative PWR fuel assembly types,” Jahrestagung Kerntechnik, Berlin, Germany, 17.19. Mai, 2011.