Transactions of the Korean Society of Mechanical Engineers B (대한기계학회논문집B)

The Korean Society of Mechanical Engineers (대한기계학회)
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1-D Two-phase Flow Investigation for External Reactor Vessel Cooling

원자로 용기 외벽냉각을 위한 1차원 이상유동 실험 및 해석

  • 김재철 (제주대학교 에너지공학과) ;
  • 박래준 (한국원자력연구원 열수력안전연구센터) ;
  • 조영로 (한국원자력연구원 열수력안전연구센터) ;
  • 김상백 (한국원자력연구원 열수력안전연구센터) ;
  • 김신 (제주대학교 에너지공학과) ;
  • 하광순 (한국원자력연구원 열수력안전연구센터)
  • Published : 2007.05.01
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Abstract

When a molten corium is relocated in a lower head of a reactor vessel, the ERVC (External Reactor Vessel Cooling) system is actuated as coolant is supplied into a reactor cavity to remove a decay heat from the molten corium during a severe accident. To achieve this severe accident mitigation strategy, the two-phase natural circulation flow in the annular gap between the external reactor vessel and the insulation should be formed sufficiently by designing the coolant inlet/outlet area and gap size adequately on the insulation device. For this reason, one-dimensional natural circulation flow tests and the simple analysis were conducted to estimate the natural circulation flow under the ERVC condition of APR1400. The experimental facility is one-dimensional and scaled down as the half height and 1/238 channel area of the APR1400 reactor vessel. The calculated circulation flow rate was similar to experimental ones within about ${\pm}$15% error bounds and depended on the form loss due to the inlet/outlet area.

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References

  1. Kymalainean, O. et al., 1997, 'In-Vessel Retention of Corium at the Loviisa Plant,' Nuclear Engineering & Design, Vol. 169, pp. 109-130 https://doi.org/10.1016/S0029-5493(96)01280-0
  2. Theofanous, T. G. et al., 1995, 'In-Vessel Coolability and Retention of a Core Melt,' DOE/ID-10460
  3. Theofanous, T. G., Yuen, W. W., Angelini, S., Sienicki, J. J,. Freeman, K., Chen, X. and Salmassi, T., 1999, 'Lower Head Integrity under Steam Explosion Loads,' Nuclear Engineering and Design, Vol. 189, pp. 7-57 https://doi.org/10.1016/S0029-5493(99)00028-X
  4. Oh, S. J. et al., 2004 'In-Vessel Retention Technology Development and Use for Advanced PWR Design in the USA and Korea,' KHNP, KHNP/TR.01NC05.C2004.EN3
  5. Jeong, Y. H., Chang, S. H. and Baek, W. P., 2003, 'CHF Experiments on the Reactor Vessel Wall using 2-D Slice Test Section,' NURETH-10, Seoul, Korea, October 5-9
  6. Rouge, S., Do, I. and Geffraye, G., 1998, 'Reactor Vessel External Cooling for Corium Retention SULTAN Experimental Program and Modeling with CATHARE Code, Workshop Proceedings on In-Vessel Core Debris Retention and Coolability,' NEA/CSNI/R(98)18, Garching, Germany, March 3-6
  7. Ha, K. S., Park, R. J., Kim, H. Y., Kim, S. B. and Kim, H. D., 2005, 'An Experimental Study on the Two-Phase Natural Circulation Flow through the Gap between Reactor Vessel and Insulation under ERVC,' KAERI Technical Report, KAERI/TR-2958/2005
  8. Cheung, F. B. and Liu, L. C., 1999, 'CHF Experiments to Support In-Vessel Retention Feasibility Study For an Evolutionary ALWR Design,' EPRI WO# 5491-01, PSU/MNE-99-263J, Feb
  9. Jong Wook Park and Dong Wook Jeong, 1997, 'An Investigation of Thermal Margin for External Reactor Vessel Cooling(ERVC) in Large Advanced Light Water Reactors(ALWR),'Proceedings of the Korean Nuclear Society Spring Meeting, Kwangju,Korea, May
  10. Ishii, M., 1977, 'One-Dimensional Drift-Flux Model and Constitutive Equations for Relative Motion Between Phases in Various Two-Phase Flow Regimes,' Argonne National Laboratory Report, ANL-77-47
  11. Neil E. Todreas and Mujid S. Kazimi, 1989, 'Nuclear System 1; Thermo Hydraulic Fundamentals,' Hemisphere Publishing Corporation, New York, p. 371, p. 379, p. 489
  12. Isao Kataoka and Ishii, M., 1987, 'Drift Flux Model for Large Diameter Pipe and New Correlation for Pool Void Fraction,' Int. J. Heat Mass Transfer https://doi.org/10.1016/0017-9310(87)90251-1
  13. Idelchik, I.E., 1996, 'Handbook of Hydraulic Resistance, 3rd Edition,' Begell House, Inc., New York, p. 171, p. 319, p. 688