Skip to main content
SpringerLink
Log in

Nano-mechanical property assessment of a neutron-irradiated HT-9 steel cladding and a fuel-cladding chemical interaction region of a uranium–10 wt% zirconium nuclear fuel

  • Original Paper
  • Published:
MRS Advances Aims and scope Submit manuscript

Abstract

Nano-indentation was used in conjunction with electron microscopy to characterize the microstructure and mechanical properties of a neutron-irradiated HT-9 cladding adjoined to uranium–10 wt% zirconium, a fast reactor fuel. Electron microscopy revealed that the neutron-irradiated cladding can be classified into three localities: HT-9 edge, HT-9 + fission products, and the fuel-cladding chemical interaction (FCCI) region. The three localities possessed differing nano-hardness values and consequently resulted in differing calculated yield stresses. Following irradiation, the nano-hardness and yield stresses increased in the HT-9 edge. The HT-9 + fission products locality also increased in nano-hardness and yield stresses with average values similar to the HT-9 edge, but with larger standard deviations due to the diffusion of fission products along grain boundaries. The FCCI locality had the largest and most variable nano-hardness and yield stresses in comparison to the other regions. A yield stress comparison of the HT-9 edge and the literature of neutron-irradiated HT-9 is discussed.

Graphical abstract

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

Access this article

Log in via an institution

Subscribe and save

Springer+ Basic
$34.99 /Month
  • Get 10 units per month
  • Download Article/Chapter or eBook
  • 1 Unit = 1 Article or 1 Chapter
  • Cancel anytime
Subscribe now

Buy Now

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

Instant access to the full article PDF.

Institutional subscriptions

Fig. 1
Fig. 2

Similar content being viewed by others

Data availability

The data required to reproduce these findings will be made available upon a reasonable request.

References

  1. O. Anderoglu, J. Van Den Bosch, P. Hosemann, E. Stergar, B.H. Sencer, D. Bhattacharyya, R. Dickerson, P. Dickerson, M. Hartl, S.A. Maloy, Phase stability of an HT-9 duct irradiated in FFTF. J. Nucl. Mater. 430(1–3), 194 (2012)

    Article  CAS  Google Scholar 

  2. C. Zheng, E.R. Reese, K.G. Field, T. Liu, E.A. Marquis, S.A. Maloy, D. Kaoumi, Microstructure response of ferritic/martensitic steel HT9 after neutron irradiation: effect of temperature. J. Nucl. Mater. 528, 151845 (2020)

    Article  CAS  Google Scholar 

  3. F.A. Garner, M.B. Toloczko, B.H. Sencer, Comparison of swelling and irradiation creep behavior of FCC-austenitic and BCC-ferritic/martensitic alloys at high neutron exposure. J. Nucl. Mater. 276(1), 123 (2000)

    Article  CAS  Google Scholar 

  4. S.J. Zinkle, G.S. Was, Materials challenges in nuclear energy. Acta Mater. 61(3), 735 (2013)

    Article  CAS  Google Scholar 

  5. R.L. Klueh, A.T. Nelson, Cladding and duct materials for advanced nuclear recycle reactors. J. Nucl. Mater. 371(1–3), 37 (2007)

    Article  CAS  Google Scholar 

  6. T.R. Allen, J.T. Busby, R.L. Klueh, S.A. Maloy, M.B. Toloczko, Cladding and duct materials for advanced nuclear recycle reactors. JOM 60(1), 15 (2008)

    Article  CAS  Google Scholar 

  7. D.S. Gelles, Development of martensitic steels for high neutron damage applications. J. Nucl. Mater. 239(1–3), 99 (1996)

    Article  CAS  Google Scholar 

  8. Y. Chen, Irradiation effects of HT-9 martensitic steel. Nucl. Eng. Technol. 45(3), 311 (2013)

    Article  CAS  Google Scholar 

  9. J.L. Séran, A. Alamo, A. Maillard, H. Touron, J.C. Brachet, P. Dubuisson, O. Rabouille, Pre- and post-irradiation mechanical properties of ferritic-martensitic steels for fusion applications: EM10 base metal and EM10/EM10 welds. J. Nucl. Mater. 212–215(PART 1), 588 (1994)

    Article  Google Scholar 

  10. P.J. Grobner, The 885°F (475°C) embrittlement of ferritic stainless steels. Metall. Trans. 4(1), 251 (1973)

    Article  CAS  Google Scholar 

  11. M. Kangilaski, Radiation effects in structural materials. React. Mater 13(3), 124 (1970)

    Google Scholar 

  12. R.L. Klueh, Elevated temperature ferritic and martensitic steels and their application to future nuclear reactors. Int. Mater. Rev. 50(5), 287 (2005)

    Article  CAS  Google Scholar 

  13. P. Dubuisson, D. Gilbon, J.L. Séran: in Int. Conf. Evol. Microstruct. Met. Dur. Irradiat. (North-Holland, 1992), p. 20

  14. P.J. Maziasz, R.L. Klueh, J.M. Vitek, Helium effects on void formation in 9Cr-1MoVNb and 12Cr-1MoVW irradiated in HFIR. J. Nucl. Mater. 141–143(PART 2), 929 (1986)

    Article  Google Scholar 

  15. E.A. Little, D.A. Stow, Void-swelling in irons and ferritic steels II. An experimental survey of materials irradiated in a fast reactor. J. Nucl. Mater. 87(1), 25 (1979)

    Article  CAS  Google Scholar 

  16. J.J. Kai, R.L. Klueh, Microstructural analysis of neutron-irradiated martensitic steels. J. Nucl. Mater. 230(2), 116 (1996)

    Article  CAS  Google Scholar 

  17. M.H. Mathon, Y. De Carlan, G. Geoffroy, X. Averty, A. Alamo, C.H. De Novion, A SANS investigation of the irradiation-enhanced α-α′ phases separation in 7–12 Cr martensitic steels. J. Nucl. Mater. 312(2–3), 236 (2003)

    Article  CAS  Google Scholar 

  18. D. S. Gelles, Effects of irradiation on low activation ferritic alloys: A review. ASTM Special Technical Publication 1047, 113 (1990)

  19. M.L. Grossbeck, L.K. Mansur, Low-temperature irradiation creep of fusion reactor structural materials. J. Nucl. Mater. 179, 130 (1991)

    Article  Google Scholar 

  20. M. Tokiwai, M. Horie, K. Kako, M. Fujiwara, Development of new ferritic steels as cladding material for metallic fuel fast breeder reactor. J. Nucl. Mater. 204, 56 (1993)

    Article  CAS  Google Scholar 

  21. W.J. Carmack, H.M. Chichester, D.L. Porter, D.W. Wootan, Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins. J. Nucl. Mater. 473, 167 (2016)

    Article  CAS  Google Scholar 

  22. A.L. Pitner, R.B. Baker, Metal fuel test program in the FFTF. J. Nucl. Mater. 204, 124 (1993)

    Article  CAS  Google Scholar 

  23. J.M. Harp, D.L. Porter, B.D. Miller, T.L. Trowbridge, W.J. Carmack, Scanning electron microscopy examination of a fast flux test facility irradiated U-10Zr fuel cross section clad with HT-9. J. Nucl. Mater. 494, 227 (2017)

    Article  CAS  Google Scholar 

  24. W.J. Carmack, D.L. Porter, Y.I. Chang, S.L. Hayes, M.K. Meyer, D.E. Burkes, C.B. Lee, T. Mizuno, F. Delage, J. Somers, Metallic fuels for advanced reactors. J. Nucl. Mater. 392(2), 139 (2009)

    Article  CAS  Google Scholar 

  25. J. Thomas, A. F. Bengoa, S. T. Nori, R. Ren, P. Kenesei, J. Almer, J. Hunter, J. Harp, M. A. Okuniewski, The application of synchrotron micro-computed tomography to characterize the three-dimensional microstructure in irradiated nuclear fuel. J. Nucl. Mater. 537, 152161 (2020)

    Article  CAS  Google Scholar 

  26. T. Milot, Establishing Correlations for Predicting Tensile Properties Based on the Shear Punch Test and Vickers Microhardness Data (M.S. Thesis, University of California, Santa Barbara, 2013)

  27. J.T. Busby, M.C. Hash, G.S. Was, The relationship between hardness and yield stress in irradiated austenitic and ferritic steels. J. Nucl. Mater. 336(2–3), 267 (2005)

    Article  CAS  Google Scholar 

  28. A. C. Fischer-Cripps: in Nanoindentation (2004), pp. 21–38

  29. D.L. Krumwiede, T. Yamamoto, T.A. Saleh, S.A. Maloy, G.R. Odette, P. Hosemann, Direct comparison of nanoindentation and tensile test results on reactor-irradiated materials. J. Nucl. Mater. 504, 135 (2018)

    Article  CAS  Google Scholar 

  30. D.L. Krumwiede, Correlation of Nanohardness to Bulk Mechanical Tensile and Shear Properties through Direct Characterization and Comparison of Neutron-Irradiated Steels (Ph.D. Dissertation, University of California, Berkeley, 2018)

  31. A. Sarkar, A.H. Alsabbagh, K.L. Murty, Investigation of microstructure and mechanical properties of low dose neutron irradiated HT-9 Steel. Ann. Nucl. Energy 65, 91 (2014)

    Article  CAS  Google Scholar 

  32. R.L. Klueh, N. Hashimoto, M.A. Sokolov, K. Shiba, S. Jitsukawa, Mechanical properties of neutron-irradiated nickel-containing martensitic steels: I. Experimental study. J. Nucl. Mater. 357(1–3), 156 (2006)

    Article  CAS  Google Scholar 

  33. S.A. Maloy, M.B. Toloczko, K.J. McClellan, T. Romero, Y. Kohno, F.A. Garner, R.J. Kurtz, A. Kimura, The effects of fast reactor irradiation conditions on the tensile properties of two ferritic/martensitic steels. J. Nucl. Mater. 356(1–3), 62 (2006)

    Article  CAS  Google Scholar 

  34. A. Almazouzi, E. Lucon, Mechanical behavior of neutron irradiated high Cr ferritic-martensitic steels. TMS Lett. 2(3), 73–74 (2005)

    CAS  Google Scholar 

  35. A.F. Rowcliffe, J.P. Robertson, R.L. Klueh, K. Shiba, D.J. Alexander, M.L. Grossbeck, S. Jitsukawa, Fracture toughness and tensile behavior of ferritic–martensitic steels irradiated at low temperatures. J. Nucl. Mater. 258, 1275 (1998)

    Article  Google Scholar 

  36. S.A. Maloy, T. Romero, M.R. James, Y. Dai, Tensile testing of EP-823 and HT-9 after irradiation in STIP II. J. Nucl. Mater. 356(1–3), 56 (2006)

    Article  CAS  Google Scholar 

  37. P. Hosemann, J.G. Swadener, D. Kiener, G.S. Was, S.A. Maloy, N. Li, An exploratory study to determine applicability of nano-hardness and micro-compression measurements for yield stress estimation. J. Nucl. Mater. 375(1), 135 (2008)

    Article  CAS  Google Scholar 

Download references

Acknowledgements

This work was supported by the US Department of Energy, Office of Nuclear Energy under DOE Idaho Operations Office Contract DE-AC07-051D14517 as part of a Nuclear Science User Facilities experiment. SEM and nano-indentation were carried out at the Irradiated Materials Characterization Laboratory at Idaho National Laboratory. This research was also partially supported by the US Nuclear Regulatory Commission through a Faculty Development Grant (M. A. Okuniewski) and fellowship (J. Thomas).

Author information

Authors and Affiliations

  1. School of Materials Engineering, Purdue University, West Lafayette, IN, 47907, USA

    Jonova Thomas & Maria A. Okuniewski

  2. Idaho National Laboratory, P.O. Box – 1625, Idaho Falls, ID, 83415, USA

    Fei Teng & Daniel Murray

Authors
  1. Jonova Thomas
    View author publications

    You can also search for this author in PubMed Google Scholar

  2. Fei Teng
    View author publications

    You can also search for this author in PubMed Google Scholar

  3. Daniel Murray
    View author publications

    You can also search for this author in PubMed Google Scholar

  4. Maria A. Okuniewski
    View author publications

    You can also search for this author in PubMed Google Scholar

Corresponding author

Correspondence to Maria A. Okuniewski.

Ethics declarations

Conflict of interest

The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper.

Rights and permissions

Reprints and permissions

About this article

Check for updates. Verify currency and authenticity via CrossMark

Cite this article

Thomas, J., Teng, F., Murray, D. et al. Nano-mechanical property assessment of a neutron-irradiated HT-9 steel cladding and a fuel-cladding chemical interaction region of a uranium–10 wt% zirconium nuclear fuel. MRS Advances 6, 1048–1053 (2021). https://doi.org/10.1557/s43580-021-00179-x

Download citation

  • Received:

  • Accepted:

  • Published:

  • Issue Date:

  • DOI: https://doi.org/10.1557/s43580-021-00179-x

Search

Navigation