Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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INVESTIGATIONS ON THE RESOLUTION OF SEVERE ACCIDENT ISSUES FOR KOREAN NUCLEAR POWER PLANTS

  • Published : 2009.06.30
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Abstract

Under the government supported long-term nuclear R&D program, the severe accident research program at KAERI is directed to investigate unresolved severe accident issues such as core debris coolability, steam explosions, and hydrogen combustion both experimentally and numerically. Extensive studies have been performed to evaluate the in-vessel retention of core debris through external reactor vessel cooling concept for APR1400 as a severe accident management strategy. Additionally, an improvement of the insulator design outside the vessel was investigated. To address steam explosions, a series of experiments using a prototypic material was performed in the TROI facility. Major parameters such as material composition and void fraction as well as the relevant physics affecting the energetics of steam explosions were investigated. For hydrogen control in Korean nuclear power plants, evaluation of the hydrogen concentration and the possibility of deflagration-to-detonation transition occurrence in the containment using three-dimensional analysis code, GASFLOW, were performed. Finally, the integrated severe accident analysis code, MIDAS, has been developed for domestication based on MELCOR. The data transfer scheme using pointers was restructured with the modules and the derived-type direct variables using FORTRAN90. New models were implemented to extend the capability of MIDAS.

Keywords

References

  1. K.J. Yoo, et al., “A study on the establishment of severe accident experiment facility,” KAERI/RR-1316/93, KAERI (1993)
  2. H.D. Kim, et al., “Experimental and Analytical Research on Severe Accident Phenomena: Research on the Containment Risk Assessment,” KAERI/RR-2227/2001, KAERI (2002)
  3. S.B. Kim, et al., “Experimental and Analytical Research on Severe Accident Phenomena: Experimental and Analytical Research on Severe Accident Phenomena,” KAERI/RR-2229/2001, KAERI (2002)
  4. T.G. Theofanous, et al., “In-vessel coolability and retention of a core melt” DOE/ID-10460, University of California (1995)
  5. V. G. Asmolov et al., “Study of Carbon and Separation Impact on Suboxidised Corium Stratification,” OECD MASCA Project Report, MP-TR-2 (2001)
  6. A.W. Cronenberg and R. Benz Vapor Explosion Phenomena with Respect to Nuclear Reactor Safety Assessment, USNRC/CR.0245-TREE-1242 (1978)
  7. G. Berthoud, Vapor Explosions, Annual Review of Fluid Mechanics, Vol. 32, pp. 573-611 (2000) https://doi.org/10.1146/annurev.fluid.32.1.573
  8. T. P. Speis and S. Basu, Fuel-coolant interaction phenomena in reactor safety: Current understanding and future research needs, Proceedings of final program of CSNI specialist meeting on fuel coolant interactions, 6-7, JAERI-Tokai, Japan (1997)
  9. Korea Electric Power Research Institute, Workshop on Invessel retention for the KNGR severe accident management (1998)
  10. D. Magallon, K-H Bang, G. Berthoud, M. Buerger, M.L. Corradini, H. Jacobs, R. Meignen, O. Melikhov, M. Naitoh, K. Moriyama, R. Sairanen, J. H. Song, N. Suh, and Theofanous, “SERENA (Steam Explosion Resolution for Nuclear Applications) Final Report,” OECD/NEA (2006)
  11. J. H. Song, I. K. Park, Y. S. Shin, J. H. Kim, S. W. Hong, B. T. Min, H. D. Kim, “Fuel Coolant Interaction Experiments in TROI Using a UO2/ZrO2 Mixture,” Nuclear Engineering and Design, 222, 1 (2003) https://doi.org/10.1016/S0029-5493(02)00388-6
  12. “Three Mile Island-Unit 2 Accident,” Nuclear Safety Analysis Center Report (1980)
  13. J. C. Cummings, A. L. Camp, M. P. Sherman, M. J. Wester, D. Tomakso, R. K. Byers, and B. W. Burnham, “Review of the Grand Gulf Hydrogen Igniter System,” SAND82- 0218, NUREG0CR-2530, Sandia National Laboratories (1983)
  14. G. Azarian, H. M. Kursawe, M. Nie, M. Fischser, J. Eyink, and R. H. Stoudt, “EPR Severe Accidents Threats and Mitigation,” Proc. Int. Congress Advances in Nuclear Power Plants (ICAPP'04), Pittsburgh, Pennsylvania, June 13-17, 2004, American Nuclear Society, (2004)
  15. B. C. Lee, J. S. Cho, G. C. Park, and C. H. Chung, “Development and Application of Two- Dimensional Hydrogen Mixing Model in Containment Subcompartment Under Severe Accidents,” J. Korean Nucl. Soc., 29, 2 (1997)
  16. OECD/NEA, “In-vessel core degradation code validation matrix,” OECD/CSNI/R (2000)21
  17. OECD/NEA, “Progress made in the last fifteen years through analyses of the TMI-2 accident performed in member countries,” NEA/CSNI/R (2005)1
  18. S.B. Kim, et al., “Optimization of the severe accident management strategy for domestic plants and validation experiments, “ KAERI/RR-2528/2004, KAERI (2005)
  19. S.B. Kim, et al., “Optimization of the severe accident management strategy for domestic plants and validation experiments, “ KAERI/RR-2786/2006, KAERI (2005)
  20. R.J. Park, et al., “Detailed analysis of the late-phase melt conditions for the evaluation of an in-vessel corium retention,” Nuclear Technology, 154, No.3 (2006)
  21. R.J. Park, S.B. Kim, and H.D. Kim, “Evaluation of the RCS depressurization strategy for the high pressure sequences by using SCDAP/RELAP5,”Annals of Nuclear Energy, 35, pp. 150-157 (2008) https://doi.org/10.1016/j.anucene.2007.07.004
  22. R.J. Park, S.B. Kim, and H.D. Kim, “Detailed analysis of in-vessel melt progression in the loss of coolant accident of OPR1000,” ICAPP '06, Reno, USA, (2006)
  23. R.J. Park, S.W. Hong, S.B. Kim, and H.D. Kim, “Detailed Evaluation of In-vessel sever accident management strategy for the SBLOCA by using SCDAP/RELAP5,” submitted to Nuclear Engineering and Technology (2008)
  24. J.W. Park, et al., “Assessment of in-vessel core debris coolability for the APR1400 design”, KHNP (2001)
  25. R.J. Park, K.S. Ha, S.B. Kim, and H.D. Kim, “Two phase natural circulation flow in the reactor cavity under an external vessel cooling during a severe accident,” Nuclear Engineering and Design, 236, pp. 2424-2430 (2006) https://doi.org/10.1016/j.nucengdes.2006.03.006
  26. D. Magallon, et al., “European expert network for the reduction of uncertainties in severe accident safety issues (EURSAFE),” Nuclear Engineering and Design 235, pp. 309-346 (2005) https://doi.org/10.1016/j.nucengdes.2004.08.042
  27. S. W. Hong, B. T. Min, J. H. Song, and H. D. Kim, "Application of Cold Crucible for Melting of UO2/ZrO2 Mixture," Materials Science and Engineering, A357, 297 (2003) https://doi.org/10.1016/S0921-5093(03)00248-X
  28. J. H. Song, B. T. Min, F. Defoort, “The Effect of Material Composition on the Strength of a Steam Explosion,” Heat Transfer Engineering, 29, Issue 8 , pp. 740-747 (2008) https://doi.org/10.1080/01457630801981754
  29. M.F.Young, “IFCI: An Integrated Code for Calculation of all Phases of Fuel-Coolant Interaction,” NUREG/CR-5084 (1987)
  30. C.C. Chu and M.L. Corradini, “One-Dimensional Transient Model for Fuel-Coolant Interaction Analysis,” Nucl. Sci. Eng. 101, pp. 48-71 (1989) https://doi.org/10.13182/NSE89-A23594
  31. W.W.Yuen and T.G.Theofanous, “The prediction of 2D thermal detonations and resulting damage potential,” Nuclear Engineering and Design, 155, pp.289-309 (1995) https://doi.org/10.1016/0029-5493(94)00878-3
  32. I.K.Park, G.C.Park, and K.H.Bang, “Multiphase Flow of Molten Material-Vapor-Liquid Mixtures in Thermal Nonequilibrium”, KSME International Journal, 14, No.5, pp.553-561 (2000) https://doi.org/10.1007/BF03185658
  33. Berthoud G, and Brayer C, “First vapor explosion calculations performed with the MC3D code.” Proc. Comm. Saf. Nucl. Install. Spec. Meet. FCI, Tokai-mura, Japan, Vol 1, pp. 391-409 (1997)
  34. Kiyofumi Moriyama, Hideo Nakamura, Yu Maruyama, “Analytical tool development for coarse break-up of a molten jet in a deep water pool,” Nuclear Engineering and Design, 236, 19-21, pp. 2010-2025 (2006) https://doi.org/10.1016/j.nucengdes.2006.03.027
  35. J. H. Song, J. H. Kim, S. W. Hong, B. T. Bin, H. D. Kim, “The effect of corium composition and interaction vessel geometry on the prototypic steam explosion,” Annals of Nuclear Energy (33), pp. 1437-1451 (2006) https://doi.org/10.1016/j.anucene.2006.09.005
  36. D. Magallon, I. Huhtiniemi, H. Hohmann, “Lessons learnt from FARO/TERMOS corium melt quenching experiments,” Nuclear Engineering and Design, 189, Issues 1-3, pp. 223-238 (1999) https://doi.org/10.1016/S0029-5493(98)00274-X
  37. D. Magallon, I. Huhtiniemi, “Corium melt quenching tests at low pressure and subcooled water in FARO,” Nuclear Engineering and Design, 204, Issues 1-3, pp. 369-376 (2001) https://doi.org/10.1016/S0029-5493(00)00318-6
  38. I. Huhtiniemi, D. Magallon, Insight into steam explosions with corium melts in KROTOS, Nuclear Engineering and Design, 204, Issues 1-3, pp. 391-400 (2001) https://doi.org/10.1016/S0029-5493(00)00319-8
  39. I. Huhtiniemi, D. Magallon, H. Hohmann, Results of recent KROTOS FCI tests: alumina versus corium melts, Nuclear Engineering and Design, 189, Issues 1-3, pp. 379-389 (1999) https://doi.org/10.1016/S0029-5493(98)00269-6
  40. M. Kato, H. Nagasaka, Y. Vasulyev, A. Kolodeshnikov, V. Zhdanov, “COTELS Project(2): Fuel Coolant Interaction Tests under Ex-vessel conditions, Proceedings of the OECD Workshop on Ex-Vessel Debris Coolability, Karlsruhe,” pp. 293-300 (1988)
  41. V. Strizhov, “Overview of the Progress in the OECD MASCA Project”, CSARP Meeting, Albuquerque, NM, USA, (2005)
  42. D. H. Cho, D. R. Armstrong, W. H., Gunther, “Experiments on Interactions between Zriconium-Containing Melt and Water,” NUREG/CR-5372, USNRC (1998)
  43. L. C. Witte, T. J. Vyas, A. A. Gelabert, “Heat Transfer and Fragmentation during Molten-Metal/Water Interactions,” Journal of Heat Transfer, pp.521-527(1973) https://doi.org/10.1115/1.3450100
  44. P. Royl, H. Rochholz, W. Breitung, J.R. Travis, G. Necker, “Analysis of steam and hydrogen distributions with PAR mitigation in NPP containments,” Nuclear Engineering and Design 202, pp. 231-248 (2000) https://doi.org/10.1016/S0029-5493(00)00332-0
  45. T.R. Travis, J.W. Spore, P. Royl, K.L. Lam, T.L. Wilson, C. Muler, G.A. Necker, B.D. Nichols, R. Redlinger, E.D. Hughes, H. Wilkening, W. Baumann, G.F. Niederauer, “GASFLOW: A computational fluid dynamics code for gases, aerosols, and combustion,” LA-13357-M, FZKA-5994, Los Alamos National Laboratory (1998)
  46. M. Houkema, E.M.J. Komen, N.B. Siccama, S.M. Willemsen, “CFD analyses of steam and hydrogen distribution in a nuclear power plant,” NURETH-10, Seoul, Korea, 2003
  47. T. Kanzleiter, Versuche zur Wirksamkeit von Wasserstroff-Gegenmassnahmen in einer Mehrraum-Containment-Geometrie, Band 1 und Band 2, Battelle Frankfurt abschlussbericht BIeV-R67.036-01 und 02 (1991)
  48. KEPCO, “Severe Accident Phenomenology and Containment Performance for the APR1400,” APR1400-SSAR (2001)
  49. S.B. Dorofeev, M. Kutznetsov, V. Alekseev, A.A. Efimenko, W. Breitung, “Evaluation of limits for effective flame acceleration in hydrogen mixtures,” Report IAE-6150/3, Kurchatov Institute Moscow (1999)
  50. W. Breitung, S.B. Dorofeev, “Criteria for Deflagration-to- Detonation Transition (DDT) in Nuclear containment Analysis,” SMIRT-15 Post-Conference Seminar on Containment of Nuclear Reactor, Seoul, Korea, (1999)
  51. S. W. Hong, Y. S. SHIN, J. H. Song, and S. H. Chang, “Performance Test of the Quenching Meshes for hydrogen Control,” Journal of Nuclear science and Technology, 40, No. 10, pp. 814-819 (2003) https://doi.org/10.3327/jnst.40.814
  52. S. W. Hong, J. H. Song, H. D. Kim and S. H. Chang, “The Applicability of a quenching mesh as a hydrogen flame arrester in Nuclear Power plants,” Nuclear Technology, 153, 89 (2006) https://doi.org/10.13182/NT06-A3691
  53. S. W. Hong, J. H. Song and H. D. Kim, “A Performance Experiment of the Quenching Mesh as a Hydrogen Control at High Flame Velocity,” Proceedings of ICAPP '04, USA, Paper 4016 (2004)
  54. R. O. Gaunt, et al., “MELCOR Computer Code Manuals,” Vol.2: Reference Manuals, Version 1.8.5, Vol. 2, Rev. 2, SAND2000-2417/2 (2000)
  55. S. H. Park, et al., “An Integration of the Restructured MELCOR for the MIDAS Computer Code,” Proceedings of International Conference on Nuclear Engineering (ICONE 14), (2006)
  56. Y. M. Song, et al., “Development of a Computer Program for Automatic variable Conversion in MELCOR Code,” Proceedings of the Korean Nuclear Autumn Meeting (2000)
  57. S. H. Park, et al., 'An Improvement of the MIDAS Computer Code for a Severe Accident Simulation (Point Kinetics),' 15th annual WIN (Women in Nuclear) Global Meeting, Indonesia (2007)
  58. J. H. Park, S. D. Kim, D. H. Kim, “The Model Development of the Gap Cooling Phenomena and Its Application to LAVA-4 Test,” KAERI/TR-1843/2001 (2001)
  59. D. H. Kim, et al., “Experimental and Analytical Research on Severe Accident Phenomena,” Development of Integrated Computer Code for Analysis of Risk Reduction Strategy, KAERI/RR-2216/2001 (2001)
  60. S. B. Kim, et al., “A parametric Study of Geometric Effect on the Debris Dispersal from a Reactor Cavity during High Pressure Melt Ejection,” Int. Communications in Heat and mass Transfer, Vol. 22, No. 1, pp.25-34 (1995) https://doi.org/10.1016/0735-1933(94)00049-Q
  61. I. K. Park, D. H. Kim, J. H. Song., “Steam Explosion Module Development for the MELCOR Code Using TEXAS-V,” Journal of the Korean Nuclear Society, Vol. 35, Number 4, pp. 286-298, August (2003)
  62. H.D. Kim, S.B. Kim, “Overview of Severe Accident Research and SONATA-IV Phase-1 Experiment at KAERI,” CSARP, Bethesda, MD, USA, (1998)

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