Nuclear Engineering and Technology

  • 제49권5호
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  • Pages.979-988
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  • 2017
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  • 1738-5733(pISSN)
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  • 2234-358X(eISSN)
한국원자력학회 (Korean Nuclear Society)
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Effect of emergency core cooling system flow reduction on channel temperature during recirculation phase of large break loss-of-coolant accident at Wolsong unit 1

  • 투고 : 2016.04.19
  • 심사 : 2017.03.31
  • 발행 : 2017.08.25
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초록

The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

키워드

참고문헌

  1. International Atomic Energy Agency, Clogged Pump Suction Strainers in the Wet Well Pool. 1992 [Internet], [cited 2016 Apr 19]. Available from: https://www-news.iaea.org.
  2. USNRC, Knowledge base for the effect of debris on pressurized water reactor emergency core cooling sump performance, NUREG/CR-6808 (2003) 1-1-1-11.
  3. OECD/NEA, Debris Impact on Emergency Coolant Recirculation, Workshop Proceedings, Feb. 25-27, 2004. Albuquerque, USA.
  4. OECD/NEA, Proceedings of the CNRA/CSNI International Workshop on Taking Account of Feedback on Sump Clogging, NEA/CSNI/R, 2009, 14, January, 2010.
  5. OECD/NEA, Update Knowledge Base for Long-term Core Cooling Reliability, NEA/CSNI/R(2013)12, November, 2013.
  6. G. Zigler, J. Brideau, D.V. Rao, C. Shaffer, E. Souto, W. Thornas, Parametric study of the potential for BWR ECCS strainer blockage due to LOCA-generated debris, NUREG/CR-6224 (1995) 1-1-2-7.
  7. B.G. Park, J.W. Park, C.H. Kim, Experimental study of debris head loss through a pressurized water reactor recirculation sump screen after LOCA, Nucl. Eng. Des. 241 (2011) 2462-2469. https://doi.org/10.1016/j.nucengdes.2011.04.013
  8. E. Krepper, G. Glover, A. Grahn, F.P. Weiss, S. Alt, R. Hampel, W. Kastner, A. Seeliger, CFD-modeling of insulation debris transport phenomena in water flow, Nucl. Eng. Des. 240 (2010) 2357-2364. https://doi.org/10.1016/j.nucengdes.2009.11.012
  9. S. Lee, Y.A. Hassan, S.S. Abdulsattar, R. Vaghetto, Experimental study of head loss through an LOCA-generated fibrous debris bed deposited on a sump strainer for Generic Safety Issue 191, Prog. Nucl. Energy 74 (2014) 166-175. https://doi.org/10.1016/j.pnucene.2014.02.019
  10. S. Rouaix, C. Laurent, Y. Armand, J.-M. Mattei, M. Liska, D. Galuskova, Y. Vicena, B. Soltesz, Precipitate formation contributing to sump screens clogging of a nuclear power plant during an accident, Chem. Eng. Res. Des. 86 (2008) 633-639. https://doi.org/10.1016/j.cherd.2008.03.016
  11. C.B. Bahn, Chemical effects on PWR sump strainer blockage after a loss-ofcoolant accident: review on U.S. research efforts, Nucl. Eng. Technol. 45 (2013) 295-310. https://doi.org/10.5516/NET.07.2013.705
  12. H. Song, S.A. Par, M. Choi, J.Y. Par, M.W. Kim, J.W. Jung, Y.T. Kim, Examination of chemical and physical effects on sump screen clogging of containment materials used in Korean plants, Ann. Nucl. Energy 69 (2014) 51-56. https://doi.org/10.1016/j.anucene.2014.01.027
  13. Korea Hydro and Nuclear Power Company, Chapter 6. Safety Systems and Chapter 15. Accident Analyses in Final Safety Analysis Report of Wolsong Unit 1, 2016.
  14. Korea Hydro and Nuclear Power Company, Stress-Test Report of Wolsong Unit 1, 2013 [Internet]. Available from: http://voc.kins.re.kr/stresstest/ [in Korean].
  15. Korea Atomic Energy Research Institute, MARS code manual, in: Code structure, System models, and Solution methods, Vol. I, 2013. KAERI/TR-2812/2004.
  16. USNRC, Improvement of RELAP5/MOD3.3.3 Models for the CANDU Plant Analysis, NUREG/IA-0189, 2000.
  17. Korea Atomic Energy Research Institute, MARS code manual, in: Developmental Assessment Report, Vol. IV, 2009. KAERI/TR-3042/2005.
  18. IAEA, Comparison of Heavy Water Reactor Thermalhydraulic Code Predictions with Small Break LOCA Experimental Data, IAEA-TECDOC-1688, 2012.
  19. Korea Institute of Nuclear Safety, Audit calculation for Wolsong unit 1 large break LOCA, 2011. KINS/HR-1129.

피인용 문헌

  1. Effect of ECCS cold-leg injection angle on thermal hydraulic characteristics and core recovery during LBLOCA in a PWR vol.142, pp.None, 2017, https://doi.org/10.1016/j.pnucene.2021.104033