DOI QR코드

DOI QR Code

Variability of plant risk due to variable operator allowable time for aggressive cooldown initiation

  • Kim, Man Cheol (School of Energy Systems Engineering, Chung-Ang University) ;
  • Han, Sang Hoon (Integrated Safety Assessment Division, Korea Atomic Energy Research Institute)
  • 투고 : 2018.10.21
  • 심사 : 2019.02.22
  • 발행 : 2019.06.25

초록

Recent analysis results with realistic assumptions provide the variability of operator allowable time for the initiation of aggressive cooldown under small break loss of coolant accident or steam generator tube rupture with total failure of high pressure safety injection. We investigated how plant risk may vary depending on the variability of operators' failure probability of timely initiation of aggressive cooldown. Using a probabilistic safety assessment model of a nuclear power plant, we showed that plant risks had a linear relation with the failure probability of aggressive cooldown and could be reduced by up to 10% as aggressive cooldown is more reliably performed. For individual accident management, we found that core damage potential could be gradually reduced by up to 40.49% and 63.84% after a small break loss of coolant accident or a steam generator tube rupture, respectively. Based on the importance of timely initiation of aggressive cooldown by main control room operators within the success criteria, implications for improvement of emergency operating procedures are discussed. We recommend conducting further detailed analyses of aggressive cooldown, commensurate with its importance in reducing risks in nuclear power plants.

키워드

참고문헌

  1. J. M, W.H. Tong, Analysis of Core Damage Frequency: Surry Power Station, Unit 1, Internal Events, NUREG/CR-4550, vol 3, United States Nuclear Regulatory Commission, 1990 rev.1.
  2. R.C. Bertucio, S.R. Brown, Analysis of Core Damage Frequency: Sequoyah, Unit 1, Internal Events. NUREG/CR-4550, vol 5, United States Nuclear Regulatory Commission, 1990 rev.1.
  3. M.B. Sattison, K.W. Hall, Analysis of Core Damage Frequency: Zion, Unit 1, Internal Events, NUREG/CR-4550, vol 7, United States Nuclear Regulatory Commission, 1990 rev.1.
  4. P. Clement, T. Chataing, R. Deruaz, OECD/NEA/CSNI International Standard Problem No.27 - BETHSY Experiment 9.1B 2" Cold Leg Break without HPSI and with Delayed Ultimate Procedure Comparison Report, NEA/CSNI/R(R92)20), vols 1 and 2, OECD Nuclear Energy Agency, 1992.
  5. H. Asaka, Y. Anoda, Y. Kukita, I. Ohtsu, Secondary-side depressurization during PWR cold-leg small break LOCAs based on ROSA-V/LSTF experiments and analyses, J. Nucl. Sci. Technol. 35 (1998) 905-915. https://doi.org/10.1080/18811248.1998.9733963
  6. T.-J. Liu, Y.-K. Chan, Y.-M. Ferng, C.-Y. Chang, Experimental investigation of early initiation of primary cooldown by secondary-side depressurization in a PWR inadequate core-cooling accident, Nucl. Technol. 129 (2000) 187-200. https://doi.org/10.13182/NT00-A3056
  7. P.A. Roth, C.J. Choi, R.R. Schultz, Analysis of Two Small Break Loss-Of-Coolant Experiments in the BETHSY Facility Using RELAP5/MOD3, EGG-NE-10353, Idaho National Engineering Laboratory and EG&G Idaho, Inc, 1992.
  8. H. Kumamaru, Y. Kukita, H. Asaka, RELAP5/MOD3 code analyses of LSTF experiments on intentional primary-side depressurization following SBLOCAs with totally failed HPI, Nucl. Technol. 126 (1999) 331-339. https://doi.org/10.13182/NT99-A2978
  9. S.H. Han, S.Y. Park, W. Jung, A Separate Report for Ulchin 3&4 Level 1 PSA, Korea Atomic Energy Research Institute, 1998.
  10. H.G. Lim, J.-H. Park, S.-C. Jang, The effect of an aggressive cool-down following a refueling outage accident in which a pressurizer safety valve is stuck open, J. Kor. Nucl. Soc. 36 (2004) 497-511.
  11. S.J. Han, H.G. Lim, J.-E. Yang, An estimation of an operator's action time by using the MARS code in a small break LOCA without a HPSI for a PWR, Nucl. Eng. Des. 237 (2007) 749-760. https://doi.org/10.1016/j.nucengdes.2006.10.016
  12. J. Cho, J.H. Park, D.-S. Kim, H.-G. Lim, Quantification of LOCA core damage frequency based on thermal-hydraulics analysis, Nucl. Eng. Des. 315 (2017) 77-92. https://doi.org/10.1016/j.nucengdes.2017.02.023
  13. D.A. Fynan, J. Cho, K.-I. Ahn, Cooldown procedure success criteria map for the full break size spectrum of SBLOCA, Nucl. Eng. Des. 326 (2018) 114-132. https://doi.org/10.1016/j.nucengdes.2017.09.022
  14. S.-J. Han, J.-E. Yang, Thermal Hydraulic Analysis of a Steam Generator Tube Rupture Accident with Total Loss of High Pressure Safety Injection, KAERI/TR-2731/2004, Korea Atomic Energy Research Institute, 2004.
  15. J.S. Kim, E.J. Jeong, M.C. Kim, Simulation analysis on steam generator tube rupture with total failure of high pressure safety injection, in: 2018 Korean Nuclear Society Spring Meeting, Jeju, Korea, 2018. May 17-18.
  16. I.B. Wall, D.W. Worledge, Some perspectives on importance measures, in:Probabilistic Safety Assessment (PSA), Utah, USA, 1996.
  17. C.L. Smith, Calculating conditional core damage probabilities for nuclear power plant operations, Reliab. Eng. Syst. Saf. 59 (1998) 299-307. https://doi.org/10.1016/S0951-8320(97)00152-X
  18. M. van der Borst, H. Schoonakker, An overview of PSA importance measures, Reliab. Eng. Syst. Saf. 72 (2001) 241-245. https://doi.org/10.1016/S0951-8320(01)00007-2
  19. J.B. Fussel, How to hand-calculate system reliability and safety characteristics, IEEE T. Reliab. R-24 (1975) 169-174. https://doi.org/10.1109/TR.1975.5215142
  20. W.E. Vesely, T.C. Davis, R.S. Denning, N. Saltos, Measures of Risk Importance and Their Applications, NUREG/CR-3385, United States Nuclear Regulatory Commission, 1983.
  21. PRiME 2.1 HUN 3&4 Model. Daejeon, Korea: Korea Atomic Energy Research Institute.
  22. S.H. Han, S.W. Lee, S.-C. Jang, H.-G. Lim, J.-E. Yang, Improved features in a PSA software AIMS-PSA, in: 2010 Korean Nuclear Society Spring Meeting, Pyeongchang, Korea, 2010.
  23. W.S. Jung, S.H. Han, J. Ha, A fast BDD algorithm for large coherent fault tree analysis, Reliab. Eng. Syst. Saf. 83 (2004) 369-374. https://doi.org/10.1016/j.ress.2003.10.009
  24. Palo Verde Individual Plant Examination, 1992. April 28.
  25. J.-J. Jeong, K.S. Ha, B.D. Chung, W.J. Lee, Development of a multi-dimensional thermal-hydraulic system code, MARS 1.3.1, Ann. Nucl. Energy 26 (1999), 1161-1642.
  26. B.-D. Chung, K.-D. Kim, S.-W. Bae, J.-J. Jeong, S.W. Lee, M.-K. Hwang, C. Yoon, MARS Code Manual Volume I: Code Structure, System Models and Solution Methods, KAERI/TR-2812/2014, Korea Atomic Energy Research Institute, 2010.
  27. J.-J. Jeong, K.D. Kim, S.W. Lee, Y.J. Lee, W.J. Lee, B.D. Chung, M. Hwang, Development of the MARS Input Model for Ulchin 3/4 Transient Analyzer, KAERI/TR-2620/2003, Korea Atomic Energy Research Institute, 2003.
  28. H.-G. Lim, J.-H. Park, S.-C. Jang, The effect of an aggressive cool-down following a refueling outage accident in which a pressurizer safety valve is stuck open, J. Kor. Nucl. Soc. 36 (2004) 497-511.
  29. S.-J. Han, H.-G. Lim, J.-E. Yang, An estimation of an operator's action time by using the MARS code in a small break LOCA without a HPSI for a PWR, Nucl. Eng. Des. 237 (2007) 749-760. https://doi.org/10.1016/j.nucengdes.2006.10.016