Sensitivity Analyses for Maximum Heat Removal from Debris in the Lower Head

  • Kim, Yong-Hoon (Seoul National University) ;
  • Kune Y. Suh (Seoul National University)
  • 발행 : 2000.08.01

초록

Parametric studies were performed to assess the sensitivity in determining the maximum in-vessel heat removal capability from the core material relocated into the lower plenum of the reactor pressure vessel (RPV)during a core melt accident. A fraction of the sensible heat can be removed during the molten jet delivery from the core to the lower plenum, while the remaining sensible heat and the decay heat can be transported by rather complex mechanisms of the counter-current flow limitation (CCFL) and the critical heat flux (CHF)through the irregular, hemispherical gap that may be formed between the freezing oxidic debris and the overheated metallic RPV wall. It is shown that under the pressurized condition of 10MPa with the sensible heat loss being 50% for the reactors considered in this study, i.e. TMI-2, KORI-2 like, YGN-3&4 like and KNGR like reactors, the heat removal through the gap cooling mechanism was capable of ensuring the RPV integrity as much as 30% to 40% of the total core mass was relocated to the lower plenum. The sensitivity analysis indicated that the cooling rate of debris coupled with the sensible heat loss was a significant factor The newly proposed heat removal capability map (HRCM) clearly displays the critical factors in estimating the maximum heat removal from the debris in the lower plenum. This map can be used as a first-principle engineering tool to assess the RPV thermal integrity during a core melt accident. The predictive model also provided ith a reasonable explanation for the non-failure of the test vessel in the LAVA experiments performed at the Korea Atomic Energy Research Institute (KAERI), which apparently indicated a cooling effect of water ingression through the debris-to-vessel gap and the intra-debris pores and crevices.

키워드

참고문헌

  1. J. L. Rempe et al., 'Investigation of the Coolability of a Continuous Mass of Relocated Debris to a Water-filled Lower Plenum,' EG&G Idoho Report, EGG-RAAM-11145 (1994)
  2. J. H. Jeong et al., 'Experimental Study on CHF in a Hemispherical Narrow Gap,' OECD/CSNI Workshop on In-Vessel Core Retention and Coolability, Garching, Germany, March (1998)
  3. K. Y. Suh and R. E. Henry, 'Integral Analysis of Debris Material and Heat Transport in Reactor Vessel Lower Plenum,' Nuclear Engineering & Design, Vol. 151, No. 1, pp. 203-221, November (1994) https://doi.org/10.1016/0029-5493(94)90043-4
  4. K. Y. Suh and R. E. Henry, 'Debris Interactions in Reactor Vessel Lower Plena During A Severe Accident - I. Predictive Model,' Nuclear Engineering & Design, Vol. 166, pp. 147-163, October (1996) https://doi.org/10.1016/0029-5493(96)01269-1
  5. K. Y. Suh and R. E. Henry, 'Debris Interactions in Reactor Vessel Lower Plena During a Severe Accident: II. Integral Analysis,' Nuclear Engineering & Design, Vol. 166, pp. 165-178, October (1996) https://doi.org/10.1016/0029-5493(96)01270-8
  6. K. Y. Suh et al., 'Melt Coolability Study Within Hemispherical Vessel Lower Plenum,' Submitted for Publication in Nuclear Technology, June (2000)
  7. R. E. Henry et al., 'An Experimental Investigation of Possible In-vessel Cooling Mechanisms,' CSARP Meeting, Bethesda, MD, USA, May (1997)
  8. Y. Maruyama et al., 'In-Vessel Debris Coolability Studies in ALPHA Program,' Proceedings of the International Topical Meeting on PSA'96, Park City, UT, USA, September (1996)
  9. T. G. Theofanous et al., 'Critical Heat Flux through Curved, Downward Facing, Thick Walls,' Nuclear Engineering and Design, Vol. 151, pp. 247-258 (1994) https://doi.org/10.1016/0029-5493(94)90046-9
  10. F. B. Cheung and K. H. Haddad, 'A Hydrodynamic Critical Heat Flux Model for Saturated Pool Boiling on a Downward Facing Saturated Pool Boiling on a Downward Facing Curved Heating Surface,' Int. J. Heat Mass Transfer, Vol. 40, No. 6, pp. 1291-1302 (1997) https://doi.org/10.1016/S0017-9310(96)00208-6
  11. I. S. Hwang et al., 'In-Vessel Retention against Water Reactor Core Melting Accidents,' Submitted for Publication in Nuclear Technology, June (2000)
  12. L. A. Stickler et al., 'Calculations to Estimate the Margin to Failure in the TMI-2 Vessel,' NUREG/CR-6196, TMI V(93)EG01, EGG-273 (1994)
  13. J. R. Wolf et al., 'TMI-2 Vessel Investigation Project Integration Report,' NUREG/CR-6197, TMI V(93)EG10, EGG-2734 (1994)
  14. OECD documents Companion Sample Examination and Related Study by JAERI H. Uetsuka;F. Nagase
  15. H. Uetsuka and F. Nagase, 'Companion Sample Examination and Related Study by JAERI,' OECD documents, pp. 269-280 (1993)
  16. P. Hofmann et al., 'Reactor Core Material Interactions at Very High Temperature,' Nuclear Technology, Vol. 87, No. 1, August (1989)
  17. J. K. Hohorst, 'SCDAP/RELAP5/MOD2 Code Manual, Volume 4: MATPRO - A Library of Materials Properties for Light-Water-Reactor Accident Analysis,' NUREG/CR-5273, EGG-2555, February (1990)
  18. O. Kymalainen et al., 'Heat Flux Distribution from a Volumetrically Heated Pool with High Rayleigh Number,' Proceedings of the 6th Nuclear Reactor Thermal Hydraulics (NURETH 6), pp. 47-53, Grenoble, France, October 5-8 (1993)
  19. H. J. Richter, 'Flooding in Tubes and Annuli,' Int. J. Multiphase Flow, Vol. 7, No. 6, pp. 647-658 (1981) https://doi.org/10.1016/0301-9322(81)90036-7
  20. K. H. Kang et al., 'Experimental Investigations on In-Vessel Debris Coolability through Inherent Cooling Mechanisms,' OECD/CSNI Workshop on In-Vessel Retention and Coolability, Garching, Germany, March 3-6 (1998)
  21. K. H. Kang et al., 'Study on the Melt Relocation Process in the In-Vessel Corium Retention Experiment (LAVA Experiment),' Proceedings of the Korean Nuclear Society Spring Meeting, Kori, Korea, May 26-27 (2000)