Mechanical Properties of Oxidized Zr-1Nb-0.7Sn-0.1Fe Alloy Nuclear Cladding Tubes at High Temperatures

Eun Kyu Lee; Sun Ig Hong

doi:10.4028/www.scientific.net/AMR.557-559.1157

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HomeAdvanced Materials ResearchAdvanced Materials Research Vols. 557-559Mechanical Properties of Oxidized...

Mechanical Properties of Oxidized Zr-1Nb-0.7Sn-0.1Fe Alloy Nuclear Cladding Tubes at High Temperatures

Abstract:

The ductility loss and the failure of cladding tubes due to oxidation are considered as a threat to the safety of the nuclear. In this study, oxidation behaviors and the effect of oxide layer on the mechanical properties of Zr-1Nb-0.7Sn-0.1Fe alloy cladding tubes with a lower tin content (<1 wt. %) were investigated. The drastic increase of the oxidation rate at 700oC suggests the change in the nature of the oxide layer with increase of temperature. The acceleration of the oxidation rate for Zr-1Nb-0.7Sn-0.1Fe alloy tubes at 700oC suggests the increased volume fraction of less protective porous oxide, supported by the presence of oxide cracks even before mechanical testing. The ductility increased after oxidation at 600oC, suggesting softening due to recovery and recrystallization is predominant over the ductility loss caused by oxidation. The ductility began to decrease significantly after oxidation at 700oC and cladding tubes exhibited the brittle fracture.

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Periodical:

Advanced Materials Research (Volumes 557-559)

Pages:

1157-1160

DOI:

https://doi.org/10.4028/www.scientific.net/AMR.557-559.1157

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Online since:

July 2012

Authors:

Eun Kyu Lee, Sun Ig Hong

Keywords:

Fuel Cladding, Mechanical Property, Oxidation, Tension Test, Zirconium Alloy

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References

[1] J.A.L. Robertson: J. Nucl. Mater. Vol. 100 (1981), p.108

Google Scholar

[2] E. Tenckhoff and P.L. Rittenhouse: J. Nucl. Mater. Vol. 35 (1970), p.14

Google Scholar

[3] E.F Ibrahim, R. Choubey and J.J. Jonas: J. Nucl. Mater. Vol. 126 (1984), p.44

Google Scholar

[4] R.A. Holt: J. Nucl. Mater. Vol. 159 (1988), p.310

Google Scholar

[5] S.Y. Lee, K.T. Kim and S.I. Hong: J. Nucl. Mater. Vol. 392 (2009), p.63

Google Scholar

[6] S. Ko, S.I. Hong and K.T. Kim: J. Nucl. Mater. Vol. 404 (2010), p.154

Google Scholar

[7] J. Godlewski, in : Zirconium in the Nuclear Industry, edited by Garde/Bradley, ASTM International Publishers, West Conshohocken (1994), Vol. 10, pp.663-683.

Google Scholar

[8] A.P. Zhilyaev and J.A. Szpunar : J. Nucl. Mater. Vol. 264 (1999), p.327

Google Scholar

[9] Y. Yan, T.A. Burtseva and M.C. Billone : J. Nucl. Mater. Vol. 393 (2009), p.433

Google Scholar

[10] M. Steinbrück and M. Böttcher : J. Nucl. Mater. Vol. 414 (2011), p.276

Google Scholar