Table of Contents Author Guidelines Submit a Manuscript
Science and Technology of Nuclear Installations
Volume 2008 (2008), Article ID 239718, 13 pages
http://dx.doi.org/10.1155/2008/239718
Research Article

Development Considerations of AREVA NP Inc.'s Realistic LBLOCA Analysis Methodology

1AREVA NP Inc., 3315 Old Forest Road, Lynchburg, VA 24506, USA
2AREVA NP Inc., 2101 Horn Rapids Road, Richland, WA 99352, USA

Received 15 May 2007; Accepted 12 December 2007

Academic Editor: Cesare Frepoli

Copyright © 2008 Robert P. Martin and Larry D. O'Dell. 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. P. S. Damerell and J. W. Simons, “2D/3D program work summary report,” Tech. Rep. NUREG/IA-0126, Nuclear Regulatory Commission, Washington, DC, USA, 1993. View at Google Scholar
  2. USNRC, “Compendium of ECCS research for realistic LOCA analysis,” Tech. Rep. NUREG/CR 1230, Nuclear Regulatory Commission, Washington, DC, USA, 1988. View at Google Scholar
  3. Technical Program Group (TPG), “Quantifying reactor safety margins,” 1989, NUREG/CR-5249, EGG-255. View at Google Scholar
  4. USNRC, “Best estimate calculations of emergency core cooling system performance,” 1989, Regulatory guide 1.15. View at Google Scholar
  5. 2001, Framatome ANP Richland Report, “Realistic large break LOCA methodology for pressurized water reactors,” EMF-2103 (proprietary.
  6. H. N. Berkow, “Safety evaluation on framatome ANP topical report accident methodology for pressurized water reactors,” April 2003, EMF-2103 (P) Rev. 0, realistic large-break loss-of-coolant, letter to J. F. Malla. View at Google Scholar
  7. T. G. Theofanous, Ed., “Preface,” Nuclear Engineering and Design, vol. 119, no. 1, p. ix, 1990. View at Publisher · View at Google Scholar
  8. T. G. Theofanous, Ed., “Preface to the discussion on quantifying reactor safety margins (pp. 405–447),” Nuclear Engineering and Design, vol. 132, no. 3, p. 403, 1992. View at Publisher · View at Google Scholar
  9. G. E. Wilson, B. E. Boyackb, I. Cattonc et al., “TPG response to the foregoing letters-to-the-editor,” Nuclear Engineering and Design, vol. 132, no. 3, pp. 431–436, 1992. View at Publisher · View at Google Scholar
  10. M. Y. Young, S. M. Bajorek, M. E. Nissley, and L. E. Hochreiter, “Application of code scaling applicability and uncertainty methodology to the large break loss of coolant,” Nuclear Engineering and Design, vol. 186, no. 1-2, pp. 39–52, 1998. View at Publisher · View at Google Scholar
  11. A. Guba, M. Makai, and L. Pál, “Statistical aspects of best estimate method—I,” Reliability Engineering and System Safety, vol. 80, no. 3, pp. 217–232, 2003. View at Publisher · View at Google Scholar
  12. W. T. Nutt and G. B. Wallis, “Evaluation of nuclear safety from the outputs of computer codes in the presence of uncertainties,” Reliability Engineering and System Safety, vol. 83, no. 1, pp. 57–77, 2004. View at Publisher · View at Google Scholar
  13. Y. Orechwa, “Comments on ‘evaluation of nuclear safety from the outputs of computer codes in the presence of uncertainties’ by W.T. Nutt and G.B. Wallis,” Reliability Engineering and System Safety, vol. 87, no. 1, pp. 133–135, 2005. View at Publisher · View at Google Scholar
  14. G. B. Wallis and W. T. Nutt, “Reply to “comments on ‘evaluation of nuclear safety from the outputs of computer codes in the presence of uncertainties’ by W.T. Nutt and G. B. Wallis” by Y. Orechwa,” Reliability Engineering and System Safety, vol. 87, no. 1, pp. 137–145, 2005. View at Google Scholar
  15. G. B. Wallis, “Uncertainties and probabilities in nuclear reactor regulation,” in Proceedings of the 11th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-11), Avignon, France, October 2005.
  16. V. H. Ransom, R. J. Wagner, J. A. Trapp et al., “RELAP5/MOD2 code manual,” 1985, NUREG/CR-4312, EGG-239. View at Google Scholar
  17. L. L. Wheat, “CONTEMPT-LT a computer program for predicting containment pressure-temperature response to a loss-of-coolant-accident,” Tech. Rep. TID-4500, ANCR-1219, Aerojet Nuclear, Idaho Falls, Idaho, USA, 1975. View at Google Scholar
  18. Siemens Power Corporation Report, “RODEX3 fuel rod thermal-mechanical response evaluation model, Vol. 1, theoretical manual; Volume 2, thermal and gas release assessments,” 1996, ANF-90-145, Vol. 1&2, Suppl. 1 (proprietary. View at Google Scholar
  19. Siemens Power Corporation Report, “S-RELAP5 programmers guide,” 2001, EMF-2101, revision 2 (proprietary. View at Google Scholar
  20. 2001, Framatome ANP Richland Report, “S-RELAP5: code verification and validation,” EMF-2102 (proprietary.
  21. 2001, Framatome ANP Richland Report, “RODEX3A: theory and users manual,” EMF-1557, revision 4, (proprietary.
  22. 2001, Framatome ANP Richland Report, “Code input development guidelines for realistic large break LOCA analysis of a pressurized water reactor,” EMF-2054, revision 2, (proprietary.
  23. 2001, Framatome ANP Richland Report, “S-RELAP5 realistic large break LOCA analysis guidelines,” EMF-2058(P), revision 1 (proprietary.
  24. 2001, Framatome ANP Richland Report, “S-RELAP5 models and correlations code manual,” EMF-2100, revision 4 (proprietary.
  25. Oak Ridge National Laboratory (ORNL), “ORNL small-break LOCA heat transfer test series I: high-pressure reflood analysis,” 1981, NUREG/CR-2114, ORNL/NUREG/TM-44. View at Google Scholar
  26. Oak Ridge National Laboratory (ORNL), “Dispersed flow film boiling in rod bundle geometry: steady state heat transfer data and correlation comparisons,” 1982, NUREG/CR-2435, ORNL-582. View at Google Scholar
  27. Oak Ridge National Laboratory (ORNL), “Analysis of transient film boiling of high-pressure water in a rod bundle,” 1982, NUREG/CR-2469, ORNL/NUREG-8. View at Google Scholar
  28. Oak Ridge National Laboratory (ORNL), “Experimental investigations of bundle boiloff and reflood under high-pressure low heat flux conditions,” 1982, NUREG/CR-2455, ORNL-584. View at Google Scholar
  29. T. M. Anklam, R. J. Miller, and M. D. White, “Experimental investigations of uncovered bundle heat transfer and two-phase mixture level swell under high pressure low heat flux conditions,” Tech. Rep. NUREG/CR-2456, ORNL-5848, Oak Ridge National Laboratory, Oak Ridge, Tenn, USA, 1982. View at Google Scholar
  30. J. A. Findlay, “BWR refill-reflood program task 4.8—model qualification task plan,” 1981, NUREG/CR-1899, EPRI NP-1527, GEAP-2489. View at Google Scholar
  31. O. Nylund et al., “Hydrodynamic and heat transfer measurements on a full-scale simulated 36-rod Marviken fuel element with uniform heat flux distribution,” 1968, ASEA and AB Atomenergi, Frigg-2, R4-447/RTL-100. View at Google Scholar
  32. A. W. Bennett, G. F. Hewitt, H. A. Kearsey, and R. K. F. Keeys, “Heat transfer to steam-water mixtures flowing in uniformly heated tubes in which the critical heat flux has been exceeded,” 1967, UKAEA Research Group Report, AERE-R 537. View at Google Scholar
  33. FLECHT SEASET Program, “PWR FLECHT SEASET unblocked bundle, forced and gravity reflood task data report,” 1980, Volumes 1 and 2. NUREG/CR-1532, EPRI NP-1459, WCAP-969. View at Google Scholar
  34. F. F. Cadek, D. P. Dominicis, and R. H. Leyse, “PWR FLECHT (full length emergency cooling heat transfer). Final report,” Tech. Rep. WCAP-7665, Chicago, Ill, USA, 1971. View at Google Scholar
  35. Siemens Power, 1998, HTP reflood test characterization report, EMF-P60,149 (proprietary.
  36. Studsvik Eco & Safety AB, “The Marviken full-scale critical flow tests,” 1982, Summary Report, NUREG/CR-2671, MXC-30. View at Google Scholar
  37. G. P. Lilly and L. E. Hochreiter, “Mixing of ECC water with steam: 1/3 scale test and summary,” EPRI Report EPRI-294-2, 1975. View at Google Scholar
  38. C. L. Tien, K. S. Chung, and C. P. Liu, “Flooding in two-phase countercurrent flows,” Tech. Rep. EPRI NP-1283, Electric Power Research Institute, Chicago, Ill, USA, 1979. View at Google Scholar
  39. E. Weiss, R. A. Markley, and A. Battacharyya, “Open duct cooling-concept for the radial blanket region of a fast breeder reactor,” Nuclear Engineering and Design, vol. 16, pp. 175–386, 1971. View at Google Scholar
  40. Siemens AG UB KWU, “Upper plenum test facility,” 1986, Test No. 12 tie plate countercurrent flow testle, R515/86/1. View at Google Scholar
  41. Siemens AG UB KWU, “Upper plenum test facility,” 1988, Test No. 8 cold/hot leg flow pattern test experimental data report, U9 316/88/1. View at Google Scholar
  42. Siemens AG UB KWU, 1988, Experimental data report—UPTF—test No. 10, tie plate countercurrent flow test, U9 316/88/.
  43. Siemens AG UB KWU, 1989, Experimental data report—UPTF—test No. 6, downcomer countercurrent flow test. U9 316/88/1.
  44. Siemens AG UB KWU, 1989, Experimental data report—UPTF—test No. 7, downcomer countercurrent flow test. U9 316/89/1.
  45. Siemens AG UB KWU, 1990, Experimental data report—UPTF—test No. 29, entrainment/deentrainment test, U9 314/90/0.
  46. Iguchi et al., 1983, Data report on large scale reflood test-43—CCTF core shakedown test C2-SH2 (Run 54). JAERI-M-58-15.
  47. T. Okubo et al., 1985, Evaluation report on CCTF core-II reflood test C2-4 (Run 62)—investigation of reproducibility. JAERI-M-85-02.
  48. H. Akimoto et al., 1987, Evaluation report on CCTF core-II reflood test C2-8 (Run 67)—effect of system pressure. JAERI-M-87-00.
  49. H. Akimoto et al., 1987, Evaluation report on CCTF core-II reflood test C2-9 (Run 68)—effect of LPCI flow rate. JAERI-M-87-00.
  50. MPR Associates, 1989, Research information report of the Slab core test facility (SCTF) core II test series. MPR-111.
  51. A. Ohnuki et al., 1991, Study on ECC injection modes in reflood tests with SCTF core II Πcomparison between gravity and forced feeds. JAERI-M-91-0.
  52. B. J. Holmes, 1991, 125 comparison report. NEA/CSNI/R(91)1, AEA-TRS-104.
  53. P. G. Pressinos et al., 1979, Experimental data report for LOFT power ascension experiment L2-3, NUGEG/CR-079.
  54. J. P. Adams et al., 1982, Quick-look report on LOFT experiment L2-5, EGG-LOFT-592.
  55. J. P. Adams et al., 1983, Quick-look report on OECD LOFT experiment LP-02-6, OECD LOFT-T-3404, EG&G Idah.
  56. J. P. Adams and J. C. Birchley, 1984, Quick-look report on OECD LOFT experiment LP-LB-1, OECD LOFT-T-3504, EG&G Idah.
  57. D. L. Reeder, “LOFT system and test description (5.5 ft Nuclear Core 1 LOCEs),” 1978, NUREG/CR-TREE-120. View at Google Scholar
  58. USNRC, “Experiment data report for semiscale Mod-1 Test S-06-3 (LOFT counterpart test),” 1978, NUREG/CR-0251;TREE-112. View at Google Scholar
  59. USNRC, “Semiscale Mod-3 Test program and system description,” 1978, NUREG/CR-0239;TREE-NUREG-121. View at Google Scholar
  60. USNRC, “Experiment data report for semiscale Mod-3 blowdown heat transfer test S-07-1 (baseline test series),” 1978, NUREG/CR-0281;TREE-122. View at Google Scholar
  61. R. P. Martin and B. M. Dunn, “Application and licensing requirements of the framatome ANP RLBLOCA methodology,” in Proceedings of the International Meeting on Updates in Best Estimate Methods in Nuclear Installation Safety Analysis (BE '04), pp. 60–70, Washington, DC, USA. View at Google Scholar
  62. E. Hofer, “The GRS programme package for uncertainty and sensitivity analysis,” in Proceedings of the Seminar on Methods and Codes for Assessing the Off-Site Consequences of Nuclear Accidents, Athens, Greece, May 1990, EUR 13013, Commission of the European Communitie.
  63. S. S. Wilks, “Determination of sample sizes for setting tolerance limits,” Annals of Mathematical Statistics, vol. 12, no. 1, pp. 91–96, 1941. View at Publisher · View at Google Scholar
  64. P. N. Somerville, “Tables for obtaining non-parametric tolerance limits,” Annals of Mathematical Statistics, vol. 29, no. 2, pp. 599–601, 1958. View at Publisher · View at Google Scholar
  65. R. P. Martin and L. D. O'Dell, “Framatome ANP's realistic large break LOCA analysis methodology,” Nuclear Engineering and Design, vol. 235, no. 16, pp. 1713–1725, 2005. View at Publisher · View at Google Scholar