Skip to main content
SpringerLink
Log in

Analysis of maximum pressure in VVER1000/V446 reactor containment for LOCA and MSLB

  • Published:
Nuclear Science and Techniques Aims and scope Submit manuscript

Abstract

The highest thermal-hydraulic pressure in the containment occurs when reactor coolant in the first loop and steam in the secondary loop discharge simultaneously, and when the maximum amount of energy from reactor unit enters to containment volume. In this paper, we investigate temperature and pressure variations in the VVER1000 containment compartments owing to concurrent break in the pipelines of the primary and secondary loops. A two-phase, multicellular model is applied in the presence of non-condensable gases. Convection and conduction through the main heat structures inside the containment are also considered. The predicted results agree well with available data. Maximum values of pressure and temperature in the containment are then calculated and compared to the design values. If LOCA and MSLB occur simultaneously, the maximum pressure would exceed the design value and integrity of the containment would be threatened.

This is a preview of subscription content, log in via an institution to check access.

Access this article

Log in via an institution

Price excludes VAT (USA)
Tax calculation will be finalised during checkout.

Instant access to the full article PDF.

Institutional subscriptions

Fig. 1
Fig. 2
Fig. 3
Fig. 4
Fig. 5
Fig. 6

Similar content being viewed by others

References

  1. D. Papini, D. Grgić, A. Cammi et al., Analysis of different containment models for IRIS small break LOCA, using GOTHIC and RELAP5 codes. Nucl. Eng. Des. 241, 1152–1164 (2011). doi:10.1016/j.nucengdes.2010.06.016

    Article  Google Scholar 

  2. T. Kim, J. Park, A containment analysis for SBLOCA in the refurbished Wolsong-1 nuclear power plant. Nucl. Eng. Des. 241, 3804–3811 (2011). doi:10.1016/j.nucengdes.2011.06.040

    Article  Google Scholar 

  3. Y. Chen, Y. Yuann, L. Dai, Lungmen ABWR containment analyses during short-term main steam line break LOCA using GOTHIC. Nucl. Eng. Des. 247, 106–115 (2012). doi:10.1016/j.nucengdes.2012.02.012

    Article  Google Scholar 

  4. L.C. Dai, Y.S. Chen, Y. Yuann, Short-term pressure and temperature MSLB response analyses for large dry containment of the Maanshan nuclear power station. Nucl. Eng. Des. 280, 86–93 (2014). doi:10.1016/j.nucengdes.2014.09.007

    Article  Google Scholar 

  5. J.C. Rosa, A. Escriva, L.E. Herranz et al., Review on condensation on the containment structures. Prog. Nucl. Energy 51, 32–66 (2009). doi:10.1016/j.pnucene.2008.01.003

    Article  Google Scholar 

  6. O. Noori-Kalkhoran, A. Minuchehr, M. Rahgoshay et al., Short-term and long-term analysis of WWER-1000 containment parameters in a large break LOCA. Prog. Nucl. Energy 71, 201–212 (2014). doi:10.1016/j.pnucene.2014.03.007

    Article  Google Scholar 

  7. O. Noori-Kalkhoran, M. Rahgoshay, A. Minuchehr et al., Analysis of thermal–hydraulic parameters of WWER-1000 containment in a large break LOCA. Ann. Nucl. Energy 68, 101–111 (2014). doi:10.1016/j.anucene.2014.01.009

    Article  Google Scholar 

  8. F.C. Rahim, M. Rahgoshay, S. Mousavian, A study of large break LOCA in the AP1000 reactor containment. Prog. Nucl. Energy 54, 132–137 (2012). doi:10.1016/j.pnucene.2011.07.004

    Article  Google Scholar 

  9. F.C. Rahim, P. Yousefi, E. Aliakbari, Simulation of the AP1000 reactor containment pressurization during loss of coolant accident. Prog. Nucl. Energy 60, 129–134 (2012). doi:10.1016/j.pnucene.2012.05.009

    Article  Google Scholar 

  10. AEOI, Final Safety Analysis Report for Bushehr VVER1000 Reactor. (Atomic Energy Organization of Iran, Tehran, 2008)

  11. O.N.W. Hargrovesd, L.J. Metcalfe, L.L. Wheat et al., CONTEMPT-LT/028—A Computer Program for Predicting Containment Pressure–Temperature Response to a Loss-of-Coolant Accident (Idaho National Engineering Laboratory, Idaho, 1979)

    Google Scholar 

  12. K. Murata, D. Williams, J. Tills, Code Manual for CONTAIN 2.0 A Computer Code for Nuclear Reactor Containment Analysis (Sandia National Laboratories, Albuquerque, NM, 1997)

    Book  Google Scholar 

  13. R.G. Irby, W.D. Crouch, R.H. Bryan, Methodology for Predicting Containment Temperatures Following a Main Steam Line Break (NEB, Tennessee, 1985)

    Google Scholar 

  14. J. Tills, A. Notafrancesco, J. Phillips, Application of the MELCOR Code to Design Basis PWR Large Dry Containment Analysis (Sandia National Laboratories, New Mexico, 2009)

    Google Scholar 

  15. NRC, Final Safety Evaluation Report Related to Certification of the AP1000 Standard Design. U.S. Nuclear Regulatory Commission, Washington, DC (2004)

  16. R.O. Gauntt, R.K. Cole, C.M. Erickson et al., MELCOR Computer Code Manuals. (NUREG/CR-6119). (Sandia National Laboratories, Albuquerque, 2000)

  17. K.R. Chun, R.A. Seban, Heat transfer to evaporating liquid films. J. Heat Transf. 93, 391–396 (1971). doi:10.1115/1.3449836

    Article  Google Scholar 

  18. F.P. Incropera, D.P. DeWitt, T.L. Bergman et al., Fundamentals of Heat and Mass Transfer (Wiley, New York, 2006), p. 583

    Google Scholar 

  19. F. Kreith, Principles of Heat Transfer (Harper & Row, New York, 1973)

    Google Scholar 

  20. R. Perry, D. Green, J. Maloney, Perry’s Chemical Engineer’s Handbook (McGraw-Hill, New York, 1997)

    Google Scholar 

Download references

Acknowledgements

The authors would like to thank S. Golabi for editing this manuscript.

Author information

Authors and Affiliations

  1. Department of Energy Engineering and Physics, Amirkabir University of Technology (Tehran Polytechnic), 424 Hafez Avenue, Tehran, 15875-4413, Iran

    Sh. Sheykhi & S. Talebi

Authors
  1. Sh. Sheykhi
    View author publications

    You can also search for this author in PubMed Google Scholar

  2. S. Talebi
    View author publications

    You can also search for this author in PubMed Google Scholar

Corresponding author

Correspondence to S. Talebi.

Rights and permissions

Reprints and permissions

About this article

Check for updates. Verify currency and authenticity via CrossMark

Cite this article

Sheykhi, S., Talebi, S. Analysis of maximum pressure in VVER1000/V446 reactor containment for LOCA and MSLB. NUCL SCI TECH 28, 132 (2017). https://doi.org/10.1007/s41365-017-0288-6

Download citation

  • Received:

  • Revised:

  • Accepted:

  • Published:

  • DOI: https://doi.org/10.1007/s41365-017-0288-6

Keywords

Search

Navigation