Creep damage characterization of UNS N10003 alloy based on a numerical simulation using the Norton creep law and Kachanov–Rabotnov creep damage model

  • Xiao-Yan Wang
  • Xiao WangEmail author
  • Xiao-Chun ZhangEmail author
  • Shi-Feng Zhu


The calculation of inelastic creep damage is important for the structural integrity evaluation of the elevated temperature structure in a thorium molten salt reactor (TMSR). However, a creep damage theory model and numerical simulation method have not been proposed for the key materials (UNS N10003 alloy) in the TMSR. In this study, creep damage characterization of UNS N10003 alloy is investigated using the Norton creep law and Kachanov–Rabotnov (K–R) creep damage model. First, the creep experimental data of the UNS N10003 alloy at 650 °C were adopted to fit the material constants of the two models. Then, the creep damage behavior of the UNS N10003 alloy was analyzed and discussed under uniaxial and multi-axial stress states. The results indicated that the K–R creep damage model is more suitable for the UNS N10003 alloy than the Norton model. Finally, the numerical simulation method was developed by a user-defined UMAT subroutine and subsequently verified through a finite element analysis (FEA). The FEA results were in agreement with the theoretical solutions. This study provides an effective method for the inelastic creep damage analysis of the elevated temperature structure in the TMSR.


Thorium molten salt reactor UNS N10003 alloy Creep damage Inelastic analysis Elevated temperature structure 


  1. 1.
    X.Z. Cai, Z.M. Dai, H.J. Xu, Thorium molten salt reactor nuclear energy system. Physics 45, 578–590 (2016). (in Chinese) CrossRefGoogle Scholar
  2. 2.
    J.J. Li, Y.L. Qian, J.L. Yin et al., Large eddy simulation of unsteady flow in gas–liquid separator applied in thorium molten salt reactor. Nucl. Sci. Tech. 29, 62 (2018). CrossRefGoogle Scholar
  3. 3.
    R.M. Ji, Y. Dai, G.F. Zhu et al., Evaluation of the fraction of delayed photoneutrons for TMSR-SF1. Nucl. Sci. Tech. 28, 135 (2017). CrossRefGoogle Scholar
  4. 4.
    ASME Boiler and Pressure Vessel Code, Section III, Division 1-Subsection NH, Class 1 Components in elevated temperature, Rules for Construction of Nuclear Facility Components (2015)Google Scholar
  5. 5.
    T. Asayama, S. Takaya, Y. Nagae et al., Creep-fatigue evaluation methodologies and related issues for Japan sodium cooled fast reactor (JSFR). Procedia Eng. 55, 309–313 (2013). CrossRefGoogle Scholar
  6. 6.
    G.H. Koo, J.J. Sienicki, C.P. Tzanos et al., Creep-fatigue design evaluations including daily load following operations for the advanced burner test reactor. Nucl. Eng. Des. 239, 1750–1759 (2009). CrossRefGoogle Scholar
  7. 7.
    G.H. Koo, J.J. Sienicki, A. Moisseytsev, Preliminary structural evaluations of the STAR-LM reactor vessel and the support design. Nucl. Eng. Des. 237, 802–813 (2007). CrossRefGoogle Scholar
  8. 8.
    H.Y. Lee, J.B. Kim, H.Y. Park, Creep-fatigue damage evaluation of sodium to air heat exchanger in sodium test loop facility. Nucl. Eng. Des. 250, 308–316 (2012). CrossRefGoogle Scholar
  9. 9.
    F. Yousefpour, S.M. Hoseyni, S.M. Hoseyni et al., Creep rupture assessment for level-2 PSA of a 2-loop PWR: accounting for phenomenological uncertainties. Nucl. Sci. Tech. 28, 107 (2017). CrossRefGoogle Scholar
  10. 10.
    H.J. Yu, Structural Mechanics of Fast Reactor (Atomic Energy Press, Beijing, 2016), pp. 61–69. (in Chinese) Google Scholar
  11. 11.
    F.H. Norton, The Creep of Steel at High Temperatures (McGraw-Hill, London, 1929)Google Scholar
  12. 12.
    X.P. Mao, Q. Guo, S.Y. Zhang et al., Study on creep damage behaviors of Ni-based alloy C276. Nucl. Power Eng. 34, 86–89 (2013). (in Chinese) CrossRefGoogle Scholar
  13. 13.
    A.A. Becker, T.H. Hyde, W. Sun et al., Benchmarks for finite element analysis of creep continuum damage mechanics. Comput. Mater. Sci. 25, 34–41 (2002). CrossRefGoogle Scholar
  14. 14.
    T.H. Hyde, A.A. Becker, A. Sun et al., Finite-element creep damage analyses of P91 pipes. Int. J. Pres. Ves. Pip. 83, 853–863 (2006). CrossRefGoogle Scholar
  15. 15.
    T. Hyde, L. Xia, Prediction of creep failure in aeroengine material under multi-axial stress states. Int. J. Mech. Sci. 38, 385–403 (1996). CrossRefGoogle Scholar
  16. 16.
    Z.H. Du, D.H. Liu, Y.H. Liu, Numerical limit load analysis of 3D pressure vessel with volume defect considering creep damage behavior. Math. Probl. Eng. 6, 1–13 (2015). CrossRefGoogle Scholar
  17. 17.
    X.C. Niu, J.M. Gong, Y. Jiang et al., Creep damage prediction of the steam pipelines with high temperature and high pressure. Int. J. Pres. Ves. Pip. 86, 593–598 (2009). CrossRefGoogle Scholar
  18. 18.
    J.T. Bao, Dissertation, Nanjing University of Technology, 2006Google Scholar
  19. 19.
    J. Lemaitre, How to use damage mechanics. Nucl. Eng. Des. 80, 233–245 (1984). CrossRefGoogle Scholar
  20. 20.
    H.T. Yao, F.Z. Xuan, Z.D. Wang et al., A review of creep analysis and design under multiaxial stress states. Nucl. Eng. Des. 237, 1969–1986 (2007). CrossRefGoogle Scholar
  21. 21.
    H.T. Yao, Dissertation, East China University of Science and Technology, 2008 (in Chinese) Google Scholar
  22. 22.
    ASME Boiler and Pressure Vessel Code, Section II, Materials, Part B, Nonferrous Material Specifications, Rules for Construction of Nuclear Facility Components (2017)Google Scholar
  23. 23.
    F.A. Leckie, D.R. Hayhurst, Constitutive equations for creep rupture. Acta Metall. 25, 1059–1070 (1977). CrossRefGoogle Scholar
  24. 24.
    S.D. Tu, Principle of High Temperature Structural Integrity (Science Press, Beijing, 2003), pp. 256–289. (in Chinese) Google Scholar
  25. 25.
    M.C. Wang, Finite Element Method (Tsinghua University Press, Beijing, 2003), pp. 185–191. (in Chinese) Google Scholar
  26. 26.
    Sh Sheykhi, S. Talebi, M. Soroush et al., Thermal-hydraulic and stress analysis of AP1000 reactor containment during LOCA in dry cooling mode. Nucl. Sci. Tech. 28, 73 (2017). CrossRefGoogle Scholar
  27. 27.
    Y. Zhong, X. Yang, D. Dong et al., Numerical study of the dynamic characteristics of a single-layer graphite core in a thorium molten salt reactor. Nucl. Sci. Tech. 29, 141 (2019). CrossRefGoogle Scholar

Copyright information

© China Science Publishing & Media Ltd. (Science Press), Shanghai Institute of Applied Physics, the Chinese Academy of Sciences, Chinese Nuclear Society and Springer Nature Singapore Pte Ltd. 2019

Authors and Affiliations

  1. 1.Shanghai Institute of Applied PhysicsChinese Academy of SciencesShanghaiChina
  2. 2.University of Chinese Academy of SciencesBeijingChina

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