DOI QR코드

DOI QR Code

Identification of hydrogen flammability in steam generator compartment of OPR1000 using MELCOR and CFX codes

  • Jeon, Joongoo (Department of Nuclear Engineering, Hanyang University) ;
  • Kim, Yeon Soo (Department of Nuclear Engineering, Hanyang University) ;
  • Choi, Wonjun (Department of Nuclear Engineering, Hanyang University) ;
  • Kim, Sung Joong (Department of Nuclear Engineering, Hanyang University)
  • 투고 : 2018.11.08
  • 심사 : 2019.06.23
  • 발행 : 2019.12.25

초록

The MELCOR code useful for a plant-specific hydrogen risk analysis has inevitable limitations in prediction of a turbulent flow of a hydrogen mixture. To investigate the accuracy of the hydrogen risk analysis by the MELCOR code, results for the turbulent gas behavior at pipe rupture accident were compared with CFX results which were verified by the American National Standard Institute (ANSI) model. The postulated accident scenario was selected to be surge line failure induced by station blackout of an Optimized Power Reactor 1000 MWe (OPR1000). When the surge line failure occurred, the flow out of the surgeline was strongly turbulent, from which the MELCOR code predicted that a substantial amount of hydrogen could be released. Nevertheless, the results indicated nonflammable mixtures owing to the high steam concentration released before the failure. On the other hand, the CFX code solving the three-dimensional fluid dynamics by incorporating the turbulence closure model predicted that the flammable area continuously existed at the jet interface even in the rising hydrogen mixtures. In conclusion, this study confirmed that the MELCOR code, which has limitations in turbulence analysis, could underestimate the existence of local combustible gas at pipe rupture accident. This clear comparison between two codes can contribute to establishing a guideline for computational hydrogen risk analysis.

키워드

참고문헌

  1. T. Nishimura, H. Hoshi, A. Hotta, Current research and development activities on fission products and hydrogen risk after the accident at Fukushima Daiichi Nuclear Power Station, Nucl. Eng. Technol. 47 (2015) 1-10. https://doi.org/10.1016/j.net.2014.12.002
  2. A. Bentaib, N. Meynet, A. Bleyer, Overview on hydrogen risk research and development activities: methodology and open issues, Nucl. Eng. Technol. 47 (2015) 26-32. https://doi.org/10.1016/j.net.2014.12.001
  3. K. Vierow, Y. Liao, J. Johnson, M. Kenton, R. Gauntt, Severe accident analysis of a PWR station blackout with the MELCOR, MAAP4 and SCDAP/RELAP5 codes, Nucl. Eng. Des. 234 (2004) 129-145. https://doi.org/10.1016/j.nucengdes.2004.09.001
  4. R.O. Gauntt, et al., MELCOR Computer Code Manuals Version 1.8.6, Sandia Natl. Lab., 2005. SAND 2005-5713.
  5. B.E. Boyack, et al., MELCOR Peer Review, Los Alamos Natl. Lab., 1992. LA-12240.
  6. N.K. Kim, J. Jeon, W. Choi, S.J. Kim, Systematic hydrogen risk analysis of OPR1000 containment before RPV failure under station blackout scenario, Ann. Nucl. Energy 116 (2018) 429-438. https://doi.org/10.1016/j.anucene.2018.02.050
  7. W. Choi, S.O. Yu, S.J. Kim, Efficacy analysis of hydrogen mitigation measures of CANDU containment under LOCA scenario, Ann. Nucl. Energy 118 (2018) 122-130. https://doi.org/10.1016/j.anucene.2018.04.008
  8. H.C. Kim, N.D. Suh, J.H. Park, Hydrogen behavior in the IRWST of APR1400 following a station blackout, Nucl. Eng. Technol. 38 (2006) 195-200.
  9. J. Wang, Y. Zhang, K. Mao, Y. Huang, W. Tian, et al., MELCOR simulation of core thermal response during a station blackout initiated severe accident in China pressurized reactor (CPR1000), Prog. Nucl. Energy 81 (2015) 6-15. https://doi.org/10.1016/j.pnucene.2014.12.008
  10. J.M. Martín-Valdepe-nas, M.A. Jimenez, F. Martin-Fuertes, J.A. Fernandez, Improvements in a CFD code for analysis of hydrogen behaviour within containments, Nucl. Eng. Des. 237 (2007) 627-647. https://doi.org/10.1016/j.nucengdes.2006.09.002
  11. C. Spengler, S. Arndt, S. Beck, J. Eckel, et al., Further Development of the Computer Codes COCOSYS and ASTEC, Gesellschaft fuer Anlagen-und Reaktorsicherheit mbH, GRS, 2014. GRS-358.
  12. RELAP4/MOD5, A Computer Program for Transient Thermal-Hydraulic Analysis of Nuclear Reactors and Related Systems User's Manual, vol. 1, Idaho Nucl. Eng. Lab., 1976. ANCR-NUREG-1335.
  13. CCPS, Understanding Atmospheric Dispersion of Accidental Releases, AIChE, New York, 1995, 6-8 and 35.
  14. S. Sklavounos, F. Rigas, Validation of turbulence models in heavy gas dispersion over obstacles, J. Hazard Mater. 108 (2004) 9-20. https://doi.org/10.1016/j.jhazmat.2004.01.005
  15. ANSYS Academic Research, ANSYS CFX User Guide, Release 17.0, ANSYS, Inc, 2015.
  16. Korea Hydro, Nuclear Power Co, Shin Kori 1&2 Final Safety Analysis Report, Seoul, Korea, 2008.
  17. J. Jeon, W. Choi, N.K. Kim, S.J. Kim, Numerical investigation of in-vessel core coolability of PWR through an effective safety injection flow model using MELCOR simulation, Ann. Nucl. Energy 121 (2018) 350-360. https://doi.org/10.1016/j.anucene.2018.07.004
  18. J. Kim, S.W. Hong, S.B. Kim, H.D. Kim, 3-Dimensional analysis of the steamhydrogen behavior from a small break loss of coolant accident in the APR1400 containment, Nucl. Eng. Technol. 36 (2004) 24-35.
  19. A.M. Gomez-Torres, E. Sainz-Mejia, J.V. Xolocostli-Munguia, et al., CFD analysis of hydrogen volumetric concentrations in a hard venting containment system of MARK-2 BWR, Ann. Nucl. Energy 85 (2015) 552-565. https://doi.org/10.1016/j.anucene.2015.06.008
  20. T. Szabo, F. Kretzschmar, T. Schulenberg, Obtaining a more realistic hydrogen distribution in the containment by coupling MELCOR with GASFLOW, Nucl. Eng. Des. 269 (2014) 330-339. https://doi.org/10.1016/j.nucengdes.2013.07.009
  21. American National Standards Institute, Design Basis for Protection of Light Water Nuclear Power Plant against the Effects of Postulated Pipe Rupture, 1988. ANSI/ANS-58.2-1988.
  22. R.K. Kumar, Flammability limits of hydrogen-oxygen-diluent mixtures, J. Fire Sci. 3 (1985) 245-262. https://doi.org/10.1177/073490418500300402
  23. J. Jeon, W. Choi, S.J. Kim, A flammability limit model for hydrogen-air-diluent mixtures based on heat transfer characteristics in flame propagation, Nucl. Eng. Technol., https://doi.org/10.1016/j.net.2019.05.005
  24. T. Aldemir, A survey of dynamic methodologies for probabilistic safety assessment of nuclear power plants, Ann. Nucl. Energy 52 (2013) 113-124. https://doi.org/10.1016/j.anucene.2012.08.001
  25. M. Heitsch, R. Huhtanen, Z. Tꠓechy, et al., CFD evaluation of hydrogen risk mitigation measures in a VVER-440/213 containment, Nucl. Eng. Des. 240 (2010) 385-396. https://doi.org/10.1016/j.nucengdes.2008.07.022
  26. B.W. Marshall, Hydrogen:Air:Steam Flammability Limits and Combustion Characteristics in the FITS Vessel, Report No. SAND84-0383, Sandia National Lab., 1986.
  27. J. Kim, S.W. Hong, Analysis of hydrogen flame acceleration in APR1400 containment by coupling hydrogen distribution and combustion analysis codes, Prog. Nucl. Energy 78 (2015) 101-109. https://doi.org/10.1016/j.pnucene.2014.09.003

피인용 문헌

  1. Recent Progress in Hydrogen Flammability Prediction for the Safe Energy Systems vol.13, pp.23, 2020, https://doi.org/10.3390/en13236263
  2. Thermal hydraulic modelling of grating effect for application to 3-dimensional analysis of hydrogen behavior in NPP containment vol.380, 2021, https://doi.org/10.1016/j.nucengdes.2021.111291