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

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주제분류원자력공학

초록

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 freefield 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

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 freefield 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

참고문헌 (25)

  1. Duy-Duan Nguyen, "Efficiency of various structural modeling schemes on evaluating seismic performance and fragility of APR1400 containment building", Nuclear Engineering and Technology, 53(8), pp.2696~2707 (2021), 한국원자력학회[DOI]
  2. L Tong, "Ultimate pressure bearing capacity analysis for the prestressed concrete containment", Annals of Nuclear Energy, 121, pp.582~593 (2018), Elsevier BV[DOI]
  3. 전영선, "CONTAINMENT PERFORMANCE EVALUATION OF PRESTRESSED CONCRETE CONTAINMENT VESSELS WITH FIBER REINFORCEMENT", Nuclear Engineering and Technology, 47(7), pp.884~894 (2015), 한국원자력학회
  4. J. Yan, "Failure mechanism of a prestressed concrete containment vessel in nuclear power plant subjected to accident internal pressure", Ann. Nucl. Energy, 133, pp.610~622 (2019)
  5. David Twidale, "Sizewell ‘B’ - a one tenth scale containment model test for the UK PWR programme", Nuclear Engineering and Design, 125(1), pp.85~93 (1991), Elsevier BV[DOI]
  6. Overpressurization Test of a 1:4-scale Prestressed Concrete Containment Vessel Model, "M.F. Hessheimer", U.S. Nuclear Regulatory Commission, ~, pp.Technical Report No. NUREG (2003), CR-6810, SAND2003-0840P
  7. Containment Integrity Research at Sandia National Laboratories-An Overview, "M.F. Hessheimer", U.S. Nuclear Regulatory Commission, ~, pp.Technical Report No. NUREG (2006), CR-6906, SAND2006-2274P
  8. Shahrokh Ghavamian, "Mechanical simulations of SANDIA II tests OECD ISP 48 benchmark", Nuclear Engineering and Design, 237(12-13), pp.1406~1418 (2007), Elsevier BV[DOI]
  9. R.M. Parmar, "Over-pressure test on BARCOM pre-stressed concrete containment", Nuclear Engineering and Design, 269, pp.177~183 (2014), Elsevier BV[DOI]
  10. Ahmad Shokoohfar, "Nonlinear analysis of pre-stressed concrete containment vessel (PCCV) using the damage plasticity model", Nuclear Engineering and Design, 298, pp.41~50 (2016), Elsevier BV[DOI]
  11. Petr Bílý, "An estimation of the effect of steel liner on the ultimate bearing capacity of prestressed concrete containment", Nuclear Engineering and Design, 328, pp.197~208 (2018), Elsevier BV[DOI]
  12. Zhi Zheng, "The optimum steel fiber reinforcement for prestressed concrete containment under internal pressure", Nuclear Engineering and Technology, 54(6), pp.2156~2172 (2022), 한국원자력학회[DOI]
  13. Qianhui Pu, "Fatigue Behavior of Prestressed Concrete Beam for Straddle-Type Monorail Tracks", Applied Sciences, 8(7), pp.1136~ (2018), MDPI AG[DOI]
  14. E. Hognestad, "A Study on Combined Bending and Axial Load in Reinforced Concrete Members", pp.1951
  15. Comparison of Different Constitutive Models for Concrete in ABAQUS, "Explicit for Missile Impact Analyses", O. Martin, 2010, JRC Scientific and Technical Reports EUR 24151 EN-2010
  16. Sara Alhanaee, "Ultimate pressure capacity of nuclear reactor containment buildings under unaged and aged conditions", Nuclear Engineering and Design, 335, pp.128~139 (2018), Elsevier BV[DOI]
  17. Risk-Informed Assessment of Degraded Containment Vessels, "B.W. Spencer", U.S. Nuclear Regulatory Commission, ~, pp.Technical Report No. NUREG (2006), CR-6920, SAND2006-3772P
  18. J. Izumo, "Analytical Model for RC Panel Elements Subjected to In-Plane Forces, vol. 12", pp.155~181 (1989), Concrete Library Japan Society of Civil Engineers
  19. Standard Specifications for Concrete Structures e Materials and Constructions, "JSCE", Concrete Committee of Japan Society of Civil Engineers, ~, pp.JSCE Guideline for Concrete No. 16 (2007)
  20. "ABAQUS-6.12, ABAQUS Analysis User's Manual. ABAQUS 6.12", Dassault Systemes Simulia Corp, ~ (2018)
  21. Lei Zhou, "Fragility comparison analysis of CPR1000 PWR containment subjected to internal pressure", Nuclear Engineering and Design, 330, pp.250~264 (2018), Elsevier BV[DOI]
  22. "Regulatory Guide 1, Containment Structural Integrity Evaluation for Internal Pressure Loadings above Design Basis Pressure", U.S. Nuclear Regulatory Commission, 216~ (2010)
  23. Capacity of Steel and Concrete Containment Vessels with Corrosion Damage, "J.L. Cherry", U.S. Nuclear Regulatory Commission, ~, pp.Technical Report No. NUREG (2001), CR-6706, SAND2000-1735
  24. J. W. Hancock, "On the mechanisms of ductile failure in highstrength steels subjected to multi-axial stress states", J. Mech. Phys. Solid, 24, pp.147~169 (1976)
  25. Song Jin, "Fragility analysis of prestressed concrete containment under severe accident condition", Annals of Nuclear Energy, 131, pp.242~256 (2019), Elsevier BV[DOI]

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