Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
권호 표지
DOI QR코드

DOI QR Code

A numerical approach for assessing internal pressure capacity at liner failure in the expanded free-field of the prestressed concrete containment vessel

  • Woo-Min Cho (Department of Nuclear Engineering, Kyung Hee University) ;
  • Seong-Kug Ha (Department of Structural Systems and Site Safety Evaluation, Korea Institute of Nuclear Safety) ;
  • SaeHanSol Kang (Department of Radiation Protection and Radioactive Waste Safety, Korea Institute of Nuclear Safety) ;
  • Yoon-Suk Chang (Department of Nuclear Engineering, Kyung Hee University)
  • Received : 2023.03.03
  • Accepted : 2023.06.20
  • Published : 2023.10.25
Download PDF
⟨ Previous Next ⟩

Abstract

Since containment building is the major shielding structure to ensure safety of nuclear power plant, the structural behavior and ultimate pressure capacity of containments must be studied in depth. This paper addresses ambiguous issue of determining free-field position for liner failure by suggesting an expanded free-field region and comparing internal pressure capacities obtained by test data, conservative assumption and suggested free-field region. For this purpose, a practical approach to determine the free-field position for the evaluation of liner tearing is carried out. The maximum principal strain histories versus internal pressure capacities among different free-field positions at various azimuths and elevations are compared with those at the equipment hatch as a conservative assumption. The comparison shows that there are considerable differences in the internal pressure capacity at liner failure within the expanded free-field region compared to the vicinity of the equipment hatch. Additionally, this study proposes an approximate correlation with conservative factors by considering the expanded free-field ranges and material characteristics to determine realistic failure criteria for liner. The applicability of the proposed correlation is demonstrated by comparing the internal pressure capacities of full-scale containment buildings following liner failure criteria according to RG 1.216 and an approximate correlation.

Keywords

Acknowledgement

This work was supported by the Nuclear Safety Research Program through the Korea Foundation Of Nuclear Safety (KoFONS) using the financial resource granted by the Nuclear Safety and Security Commission (NSSC) of the Republic of Korea (No. 2106034).

References

  1. D. Nguyen, B. Thusa, H.S. Park, M.S. Azad, T.H. Lee, Efficiency of various structural modeling shemes on evaluating seismic performance and fragility of APR 1400 containment building, Nucl. Eng. Technol. 53 (8) (2021) 2696-2707. https://doi.org/10.1016/j.net.2021.02.006
  2. L. Tong, X. Zhou, X. Cao, Ultimate pressure bearing capacity analysis for the prestressed concrete containment, Ann. Nucl. Energy 121 (2018) 582-593, https://doi.org/10.1016/j.anucene.2018.08.020.
  3. Y.S. Choun, H.K. Park, Containment performance evaluation of prestressed concrete containment vessels with fiber reinforcement, Nucl. Eng. Technol. 47 (7) (2015) 884-894. https://doi.org/10.1016/j.net.2015.07.003
  4. J. Yan, Y. Lin, Z. Wang, T. Fang, J. Ma, Failure mechanism of a prestressed concrete containment vessel in nuclear power plant subjected to accident internal pressure, Ann. Nucl. Energy 133 (2019) 610-622. https://doi.org/10.1016/j.anucene.2019.07.013
  5. D. Twidale, R. Crowder, Sizewell 'B'-a one tenth scale containment model test for the UK PWR programme, Nucl. Eng. Des. 125 (1) (1991) 85-93, https://doi.org/10.1016/0029-5493(91)90008-6.
  6. M.F. Hessheimer, E.W. Klamerus, L.D. Lambert, G.S. Rightley, R.A. Dameron, Overpressurization Test of a 1:4-scale Prestressed Concrete Containment Vessel Model. Technical Report No. NUREG/CR-6810, SAND2003-0840P, U.S. Nuclear Regulatory Commission, Washington, DC, USA, 2003.
  7. M.F. Hessheimer, R.A. Dameron, Containment Integrity Research at Sandia National Laboratories-An Overview. Technical Report No. NUREG/CR-6906, SAND2006-2274P, U.S. Nuclear Regulatory Commission, Washington, DC, USA, 2006.
  8. S. Ghavamian, A. Courtois, J.L. Valfort, Mechanical simulations of SANDIA II tests OECD ISP 48 benchmark, Nucl. Eng. Des. 237 (12-13) (2007) 1406-1418, https://doi.org/10.1016/j.nucengdes.2006.10.012.
  9. R.M. Parmar, T. Singh, I. Thangamani, N. Trivedi, R.K. Singh, Over-pressure test on BARCOM pre-stressed concrete containment, Nucl. Eng. Des. 269 (2014) 177-183, https://doi.org/10.1016/j.nucengdes.2013.08.027.
  10. A. Shokoohfar, A. Rahai, Nonlinear analysis of pre-stressed concrete containment vessel (PCCV) using the damage plasticity model, Nucl. Eng. Des. 298 (2016) 41-50, https://doi.org/10.1016/j.nucengdes.2015.12.019.
  11. P. Bily, A. Kohoutkova, An estimation of the effect of steel liner on the ultimate bearing capacity of prestressed concrete containment, Nucl. Eng. Des. 328 (2018) 197-208, https://doi.org/10.1016/j.nucengdes.2017.12.026.
  12. Z. Zheng, Y. Sun, X. Pan, C. Su, J. Kong, The optimum steel fiber reinforcement for prestressed concrete containment under internal pressure, Nucl. Eng. Technol. 54 (6) (2022) 2156-2172. https://doi.org/10.1016/j.net.2021.12.015
  13. Q. Pu, H. Wang, H. Gou, Y. Bao, M. Yan, Fatigue behavior of prestressed concrete beam for straddle-type monorail tracks, Appl. Sci. 8 (7) (2018), https://doi.org/10.3390/app8071136.
  14. E. Hognestad, A Study on Combined Bending and Axial Load in Reinforced Concrete Members, University of Illinois at Urbana-Champaign, IL, Bulletin, 1951, pp. 43-46.
  15. O. Martin, Comparison of Different Constitutive Models for Concrete in ABAQUS/Explicit for Missile Impact Analyses. JRC Scientific and Technical Reports EUR 24151 EN-2010, European Commission Joint Research Centre Institute for Energy, Westerduinweg, Netherland, 2010.
  16. S. Alhanaee, Y. Yi, A. Schiffer, Ultimate pressure capacity of nuclear reactor containment building under unaged and aged conditions, Nucl. Eng. Des. 335 (2018) 128-139, https://doi.org/10.1016/j.nucengdes.2018.05.017.
  17. B.W. Spencer, J.P. Petti, D.M. Kunsman, Risk-Informed Assessment of Degraded Containment Vessels, U.S. Nuclear Regulatory Commission, Washington, DC, USA, 2006. Technical Report No. NUREG/CR-6920, SAND2006-3772P.
  18. J. Izumo, H. Shima, H. Okamura, Analytical Model for RC Panel Elements Subjected to In-Plane Forces, vol. 12, Concrete Library Japan Society of Civil Engineers, 1989, pp. 155-181.
  19. JSCE, JSCE Guideline for Concrete No. 16, Standard Specifications for Concrete Structures e Materials and Constructions, Concrete Committee of Japan Society of Civil Engineers, Japan, 2007.
  20. ABAQUS-6.12, ABAQUS Analysis User's Manual. ABAQUS 6.12, Dassault Systemes Simulia Corp., Providence, RI, USA, 2018.
  21. L. Zhou, J. Li, H. Zhong, G. Lin, Z. Li, Fragility comparison analysis of CPR 1000 PWR containment subjected to internal pressure, Nucl. Eng. Des. 330 (2018) 250-264, https://doi.org/10.1016/j.nucengdes.2018.02.005.
  22. Regulatory Guide 1, Containment Structural Integrity Evaluation for Internal Pressure Loadings above Design Basis Pressure, U.S. Nuclear Regulatory Commission, Rockville, MD, USA, 2010, 216.
  23. J.L. Cherry, J.A. Smith, Capacity of Steel and Concrete Containment Vessels with Corrosion Damage. Technical Report No. NUREG/CR-6706, SAND2000-1735, U.S. Nuclear Regulatory Commission, Washington, DC, USA, 2001.
  24. J.W. Hancock, A.C. Mackenzie, On the mechanisms of ductile failure in high-strength steels subjected to multi-axial stress states, J. Mech. Phys. Solid. 24 (1976) 147-169. https://doi.org/10.1016/0022-5096(76)90024-7
  25. S. Jin, Z. Li, T. Lan, J. Gong, Fragility analysis of prestressed concrete containment under severe accident condition, Ann. Nucl. Energy 131 (2019) 242-256, https://doi.org/10.1016/j.anucene.2019.03.034.