Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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EVALUATION OF HEAT-FLUX DISTRIBUTION AT THE INNER AND OUTER REACTOR VESSEL WALLS UNDER THE IN-VESSEL RETENTION THROUGH EXTERNAL REACTOR VESSEL COOLING CONDITION

  • JUNG, JAEHOON (Severe Accident and PHWR Safety, Korea Atomic Energy Research Institute) ;
  • AN, SANG MO (Severe Accident and PHWR Safety, Korea Atomic Energy Research Institute) ;
  • HA, KWANG SOON (Severe Accident and PHWR Safety, Korea Atomic Energy Research Institute) ;
  • KIM, HWAN YEOL (Severe Accident and PHWR Safety, Korea Atomic Energy Research Institute)
  • Received : 2014.07.29
  • Accepted : 2014.11.17
  • Published : 2015.02.25
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Abstract

Background: A numerical simulation was carried out to investigate the difference between internal and external heat-flux distributions at the reactor vessel wall under in-vessel retention through external reactor vessel cooling (IVR-ERVC). Methods: Total loss of feed water, station blackout, and large break loss of coolant accidents were selected as the severe accident scenarios, and a transient analysis using the element-birth-and-death technique was conducted to reflect the vessel erosion (vessel wall thickness change) effect. Results: It was found that the maximum heat flux at the focusing region was decreased at least 10% when considering the two-dimensional heat conduction at the reactor vessel wall. Conclusion: The results show that a higher thermal margin for the IVR-ERVC strategy can be achieved in the focusing region. In addition, sensitivity studies revealed that the heat flux and reactor vessel thickness are dominantly affected by the molten corium pool formation according to the accident scenario.

Keywords

Acknowledgement

Supported by : National Research Foundation of Korea

References

  1. J.L. Rempe, K.Y. Suh, F.B. Cheung, S.B. Kim, In-vessel retention of molten corium: lessons learned and outstanding issues, Nuclear Technology 161 (2008) 210e267. https://doi.org/10.13182/NT08-A3924
  2. F.B. Cheung, Limiting factors for external reactor vessel cooling, Nuclear Technology 152 (2005) 145e161. https://doi.org/10.13182/NT05-A3666
  3. K.S. Ha, F.B. Cheung, J.H. Song, R.J. Park, S.B. Kim, Prediction of boiling-induced natural circulation flow in engineered cooling channels, Nuclear Technology 181 (2013) 133e143. https://doi.org/10.13182/NT13-A15762
  4. T.G. Theofanous, M. Maguire, S. Angelini, T. Salmassi, The first results from the ACOPO experiment, in: Proc., the International Topical Meeting on Probabilistic Safety Assessment (PSA'96), 1996.
  5. T.N. Dinh, J.P. Tu, T. Salmassi, T.G. Theofanous, Limits of Coolability in the AP1000-related ULPU-2400 Configuration V Facility, Rep. No. CRSS-03006, Center for Risk Studies and Safety, University of California, Santa Barbara, CA, 2003.
  6. Y.H. Jeong, S.H. Chang, Critical heat flux experiments on the reactor vessel wall using 2-d slice test section, Nuclear Technology 152 (2004) 162e169.
  7. T.G. Theofanous, C. Liu, S. Additon, S. Angelini, O. Kymalainen, T. Salmassi, In-vessel coolability and retention of a core melt, Nuclear Engineering and Design 169 (1997) 1e48. https://doi.org/10.1016/S0029-5493(97)00009-5
  8. R.J. Park, J.R. Lee, K.S. Ha, H.Y. Kim, Evaluation of in-vessel corium retention through external reactor cooling for small integral reactor, Nuclear Engineering and Design 262 (2013) 571e578. https://doi.org/10.1016/j.nucengdes.2013.06.003
  9. J.L. Rempe, K.Y. Suh, F.B. Cheung, S.B. Kim, In-vessel retention strategy for high-power reactors final report, INEEL/EXT-04-02561, Idaho National Laboratory Report, January 2005, 2005.
  10. Y.A. Cengel, A.J. Ghajar, Heat and Mass Transfer, fourth ed., McGraw-Hill Science/Engineering/Math, NY, Feb. 22, 2010.
  11. R.N. Mhetre, S.G. Jadhav, Finite element analysis of welded joints, International Journal of Instrumentation, Control and Automation 1 (3e4) (2012) 116e120. ISSN: 2231-1890, http://www.interscience.in/IJICA_Vol1Iss3-4/116-120.pdf.

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