Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Burst pressure estimation of Alloy 690 axial cracked steam generator U-bend tubes using finite element damage analysis

  • Kim, Ji-Seok (Korea University, Department of Mechanical Engineering) ;
  • Kim, Yun-Jae (Korea University, Department of Mechanical Engineering) ;
  • Lee, Myeong-Woo (Korea Atomic Energy Research Institute, Innovative SMR System Development Division) ;
  • Jeon, Jun-Young (Korea Institute of Materials Science, Authorized Nuclear Inspection Department) ;
  • Kim, Jong-Sung (Sejong University, Department of Nuclear Engineering)
  • Received : 2020.02.20
  • Accepted : 2020.07.16
  • Published : 2021.02.25
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Abstract

This paper presents numerical estimation of burst pressures of axial cracked U-bend tubes, considering the U-bending process analysis. The validity of the FE simulations is confirmed by comparing with published experimental data. From parametric analyses, it is shown that existing EPRI burst pressure estimation equations for straight tubes can be conservatively used to estimate burst pressures of the U-bend tubes. This is due to the increase in yield strength during the U-bending process. The degree of conservatism would decrease with increasing the bend radius and with increasing the crack depth.

Keywords

Acknowledgement

This study was supported by Korea Institute of Energy Technology Evaluation and Planning (KETEP) (No. 20171520101650).

References

  1. S. Newberry, J. Donoghue, R. Ennis, L. Lund, A. Rubin, J. Yerokun, Indian point 2 steam generator tube failure lessons-learned report, in: TAC No. MA9163, U.S. Nuclear Regulatory Commission, Washington, DC, 2000 Oct 23.
  2. J. Allan, S. Kaufmann, E. Murphy, J. Strosnider, M. Gamborini, D. Lu, T. Sullivan, T. Pitterle, T. Esselman, A. Neff, J. Mark, J. Parry, J. Baumstark, D. Mcadam, D. Adanonis, G. Henry, Official Transcript of Proceedings United States of America Nuclear Regulatory Commission: NRC and Consolidated Edison Technical Meeting Regarding IP2 Steam Generator, U.S. Nuclear Regulatory Commission, Washington, DC, 2000 May 3.
  3. T.J. McGinty, Crack-like Indication in the U-Bend Region of a Thermally Treated Alloy 600 Steam Generator Tube, 2010 Oct 6. NRC information notice 2010-21.
  4. P.E. MacDonald, V.N. Shah, L.W. Ward, P.G. Ellison, Steam Generator Tube Failures. Idaho Falls (ID): Idaho National Engineering Lab, Div. Of Safety Programs, U.S. Nuclear Regulatory Commission, Washington, DC, 1996 Apr 1. Report No. NUREG/CR-6365, INEL-95/0383.
  5. B. Flesch, B. Cochet, Leak-before-break in steam generator tubes, Int. J. Pres. Ves. Pip. 43 (1990) 165-179. https://doi.org/10.1016/0308-0161(90)90099-4
  6. J. Baron, W. Cullen, J. Smith, L. Genutis, K. Colgan, Steam Generator Management Program: Steam Generator Degradation Specific Management Flaw Handbook, Revision 2, Cranberry Township (PA), Westinghouse Electric Company, Palo Alto, CA, 2015 Nov 03. Report No. 3002005426. Sponsored by Electric Power Research Institute.
  7. C.B. Bahn, Y.J. Oh, S. Majumdar, Ligament rupture and unstable burst behaviors of axial flaws in steam generator U-bends, Nucl. Eng. Des. 293 (17-19) (2015) 228-237. https://doi.org/10.1016/j.nucengdes.2015.06.019
  8. J.Y. Jeon, Y.J. Kim, J.W. Kim, K.H. Lee, J.S. Kim, Numerical prediction of maximum load-carrying capacity of cracked alloy 690TT steam generator tubes, J. Pressure Vessel Technol. 138 (4) (2016), 041601. https://doi.org/10.1115/1.4031746
  9. M.W. Lee, J.S. Kim, J.Y. Jeon, Y.J. Kim, J.W. Kim, J.S. Kim, Comparison of numerical predictions with experimental burst pressures of tubes with multiple surface cracks, Eng. Fract. Mech. 201 (2018) 176-195. https://doi.org/10.1016/j.engfracmech.2018.06.016
  10. J.S. Kim, M.W. Lee, Y.J. Kim, J.W. Kim, Numerical validation of burst pressure estimation equations for steam generator tubes with multiple axial surface cracks, Nucl. Eng. Technol. 51 (2) (2019) 579-587. https://doi.org/10.1016/j.net.2018.10.013
  11. K.H. Eom, J.W. Kim, Evaluation of tensile properties of alloy 690TT steam generator tube at room temperature and 343℃, Trans. Kor. Soc. Mech. Eng. A 38 (6) (2014) 655-662 [Korean]. https://doi.org/10.3795/KSME-A.2014.38.6.655
  12. Ph. Berge, H.D. Bui, J.R. Donati, D. Villard, Residual stresses in bent tubes for nuclear steam generators, Corrosion 32 (9) (1976) 357-364. https://doi.org/10.5006/0010-9312-32.9.357
  13. H.S. Nam, Y.R. Oh, Y.J. Kim, J.S. Kim, N. Miura, Application of engineering ductile tearing simulation method to CRIEPI pipe test, Eng. Fract. Mech. 153 (2016) 128-142. https://doi.org/10.1016/j.engfracmech.2015.12.012
  14. H.W. Ryu, K.D. Bae, Y.J. Kim, J.J. Han, J.S. Kim, P.J. Budden, Ductile tearing simulation of Battelle pipe test using simplified stress-modified fracture strain concept, Fatig. Fract. Eng. Mater. Struct. 39 (11) (2016) 1391-1406. https://doi.org/10.1111/ffe.12456
  15. C.S. Oh, N.H. Kim, Y.J. Kim, J.H. Baek, Y.P. Kim, W.S. Kim, A finite element ductile failure simulation method using stress-modified fracture strain model, Eng. Fract. Mech. 78 (1) (2011) 124-137. https://doi.org/10.1016/j.engfracmech.2010.10.004
  16. F.A. McClintock, A criterion of ductile fracture by the growth of holes, J. Appl. Mech. 35 (2) (1968) 363-371. https://doi.org/10.1115/1.3601204
  17. J.R. Rice, D.M. Tracey, On the ductile enlargement of voids in triaxial stress fields, J. Mech. Phys. Solid. 17 (3) (1969) 201-217. https://doi.org/10.1016/0022-5096(69)90033-7
  18. J.W. Hancock, A.C. Mackenzie, On the mechanisms of ductile failure in highstrength steels subject to multi-axial stress states, J. Phys. Mech. Solids 24 (2-3) (1976) 147-160. https://doi.org/10.1016/0022-5096(76)90024-7
  19. G.R. Jonson, W.H. Cook, Fracture characteristics of three metals subjected to various strains, strain rates, temperatures and pressures, Eng. Fract. Mech. 21 (1) (1985) 31-48. https://doi.org/10.1016/0013-7944(85)90052-9
  20. Dassault, ABAQUS Version 6.18. User's Manual, Dassault Systemes Simulia, Providence, RI, 2018.

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