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Science and Technology of Nuclear Installations
Volume 2014 (2014), Article ID 292916, 18 pages
http://dx.doi.org/10.1155/2014/292916
Research Article

Applying UPC Scaling-Up Methodology to the LSTF-PKL Counterpart Test

Technical University of Catalonia, Institute of Energy Technologies, Avenida Diagonal 647, 08028 Barcelona, Spain

Received 2 September 2013; Accepted 12 December 2013; Published 2 March 2014

Academic Editor: Eugenijus Ušpuras

Copyright © 2014 V. Martinez-Quiroga 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. “Accident analysis for nuclear power plants,” Safety Reports Series 23, IAEA, Vienna, Austria, 2002.
  2. “Development and review of plant specific emergency operating procedures,” Safety Reports Series 48, IAEA, Vienna, Austria, 2006.
  3. F. Reventos, “Thermal-hydraulic analysis tasks for ANAV NPPs in support of plant operation and control,” Science and Technology of Nuclear Installations, vol. 2008, Article ID 153858, 13 pages, 2008. View at Publisher · View at Google Scholar
  4. F. Reventós, C. Llopis, L. Batet, C. Pretel, and I. Sol, “Analysis of an actual reactor trip operating event due to a high variation of neutron flux occurring in the Vandells-II nuclear power plant,” Nuclear Engineering and Design, vol. 240, no. 10, pp. 2999–3008, 2010. View at Publisher · View at Google Scholar · View at Scopus
  5. C. Llopis, F. Reventós, L. Batet, C. Pretel, and I. Sol, “Analysis of low load transients for the Vandellòs-II NPP. Application to operation and control support,” Nuclear Engineering and Design, vol. 237, no. 18, pp. 2014–2023, 2007. View at Publisher · View at Google Scholar · View at Scopus
  6. F. Reventós, L. Batet, C. Llopis, C. Pretel, M. Salvat, and I. Sol, “Advanced qualification process of ANAV NPP integral dynamic models for supporting plant operation and control,” Nuclear Engineering and Design, vol. 237, no. 1, pp. 54–63, 2007. View at Publisher · View at Google Scholar · View at Scopus
  7. A. Cuadra, J. Gago, and F. Reventos, “Analysis of a main steam line break in ascó nuclear power plant,” Nuclear Technology, vol. 146, no. 1, pp. 41–48, 2004. View at Google Scholar
  8. V. Martinez-Quiroga and F. Reventos, “The use of system codes in scaling studies: relevant techniques for qualifying NPP nodalizations for particular scenarios,” Science and Technology of Nuclear Installations, vol. 2014, Article ID 138745, 13 pages, 2014. View at Publisher · View at Google Scholar
  9. H. Kremin, H. limprecht, R. Güneysu, and K. Umminger, “Description of the PKL III test facility,” Framatome ANP Report, 2001. View at Google Scholar
  10. The ROSA-V Group, “ROSA-V Large Scale Test Facility (LSTF) system description for the third and fourth simulated fuel assemblies,” Technical Report JAERI-Tech 2003-037, Japan Atomic Energy Agency, 2003. View at Google Scholar
  11. J. Freixa, F. Reventós, C. Pretel, L. Batet, and I. Sol, “SBLOCA with boron dilution in pressurized water reactors. Impact on operation and safety,” Nuclear Engineering and Design, vol. 239, no. 4, pp. 749–760, 2009. View at Publisher · View at Google Scholar · View at Scopus
  12. J. Freixa, SBLOCA with boron dilution in pressurized water reactors. Impact to the operation and safety [Ph.D. thesis], Universitat Politècnica de Catalunya, DFEN, 2007.
  13. J. Freixa, F. Reventós, C. Pretel, and L. Batet, “Boron transport model with physical diffusion for RELAP5,” Nuclear Technology, vol. 160, no. 2, pp. 205–215, 2007. View at Google Scholar · View at Scopus
  14. F. Reventós, J. Freixa, L. Batet et al., “An analytical comparative exercise on the OECD-SETH PKL E2.2 experiment,” Nuclear Engineering and Design, vol. 238, no. 4, pp. 1146–1154, 2008. View at Publisher · View at Google Scholar · View at Scopus
  15. “RELAP5/MOD3 Code manual. Volume II: Appendix A Input Requirements,” NUREG/CR-5535/Rev.1, ISL, Idaho, USA, January 2003.
  16. V. Martínez, F. Reventós, and C. Pretel, “Post-Test Calculation of the ROSA/LSTF Test 3-1 using RELAP5/mod3.3,” NUREG/IA-409, 2012.
  17. V. Martínez, F. Reventós, and C. Pretel, “Post-Test Calculation of the ROSA/LSTF Test 3-2 using RELAP5/mod3.3,” NUREG/IA-410, 2012.
  18. F. D'Auria and G. M. Galassi, “Scaling in nuclear reactor system thermal-hydraulics,” Nuclear Engineering and Design, vol. 240, no. 10, pp. 3267–3293, 2010. View at Publisher · View at Google Scholar · View at Scopus