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Creep Behavior of Hydrogenated Zirconium Alloys

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Abstract

Zirconium (Zr) alloys are the primary structural materials of most water reactors. Creep is considered to be one of the important degradation mechanisms of Zr alloys during reactor operating and repository conditions. Zr alloys pick up hydrogen (H2) during their service from the coolant water. Hydrogen can be present in solid solution or precipitated hydride form in Zr alloys depending upon the temperature and concentration. This study reviews the effect of hydrogen on creep behavior of Zr alloys used in the water reactors.

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References

  1. K. Linga Murty and I. Charit, Texture Development and Anisotropic Deformation of Zircaloys, Prog. Nucl. Energy, 2006, 48(4), p 325–359

    Article  Google Scholar 

  2. A.T. Motta et al., Zirconium Alloys for Supercritical Water Reactor Applications: Challenges and Possibilities, J. Nucl. Mater., 2007, 371(1–3), p 61–75

    Article  Google Scholar 

  3. S. Banerjee and P. Mukhopadhyay, Phase Transformations: Examples from Titanium and Zirconium Alloys, Vol 12, Elsevier Science, Oxford, 2010

    Google Scholar 

  4. G. Östberg, Determination of Hydride Solubility in Alpha Phase Zirconium, Zircaloy-2 and Zircaloy 4, J. Nucl. Mater., 1962, 5(2), p 208–215

    Article  Google Scholar 

  5. R. Singh et al., Terminal Solid Solubility of Hydrogen in Zr-Alloy Pressure Tube Materials, J. Alloy. Compd., 2005, 389(1), p 102–112

    Article  Google Scholar 

  6. M.P. Puls, The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Components: Delayed Hydride Cracking, Springer, New York, 2012

    Book  Google Scholar 

  7. R.N. Singh et al., Temperature Dependence of Misfit Strains of δ-Hydrides of Zirconium, J. Alloy. Compd., 2007, 436(1), p 150–154

    Article  Google Scholar 

  8. K.L. Murty, Materials Ageing and Degradation in Light Water Reactors: Mechanisms and Management, Elsevier, Amsterdam, 2013

    Book  Google Scholar 

  9. C. Chow et al., Case Histories Involving Fatigue and Fracture Mechanics, American Society for Testing and Materials, Philadelphia, 1986, p 389–416

    Google Scholar 

  10. K. Chan, A Micromechanical Model for Predicting Hydride Embrittlement in Nuclear Fuel Cladding Material, J. Nucl. Mater., 1996, 227(3), p 220–236

    Article  Google Scholar 

  11. R. Singh et al., Delayed Hydride Cracking in Zr-2.5Nb Pressure Tube Material, J. Nucl. Mater., 2002, 304(2), p 189–203

    Article  Google Scholar 

  12. R. Singh et al., Delayed Hydride Cracking in Zr-2.5Nb Pressure Tube Material: Influence of Fabrication Routes, Mater. Sci. Eng. A, 2004, 374(1), p 342–350

    Article  Google Scholar 

  13. R. Singh et al., Influence of Hydrogen Content on Impact Toughness of Zr-2.5Nb Pressure Tube Alloy, Nucl. Eng. Des., 2011, 241(7), p 2425–2436

    Article  Google Scholar 

  14. P. Efsing and K. Pettersson, The Influence of Temperature and Yield Strength on Delayed Hydride Cracking in Hydrided Zircaloy-2, ASTM Spec. Tech. Publ., 1996, 1295, p 394–403

    Google Scholar 

  15. J. Wei, Effect of Hydrogen on the Corrosion Performance of Zirconium Alloys, The University of Manchester, Manchester, 2012

    Google Scholar 

  16. K. Murty, The Internal Pressurization Creep of Zr Alloys for Spent-Fuel Dry Storage Feasibility, JOM, 2000, 52(9), p 34–38

    Article  Google Scholar 

  17. W. Daugherty and K. Murty, Application of Texture in Predicting Nuclear Fuel Cladding Creep, Nucl. Technol., 1988, 60(3), p 443–450

    Google Scholar 

  18. R.E. Einziger, Extending Dry Storage of Spent LWR Fuel for 100 Years, Argonne National Lab, Lemont, IL, 1998

    Google Scholar 

  19. M. Quecedo et al., Results of Thermal Creep Test on Highly Irradiated ZIRLO, Proceedings of the International LWR Fuel Performance Meeting, Seul, Paper, 2008

  20. R. Adamson, F. Garzarolli, and C. Patterson, In-Reactor Creep of Zirconium Alloys, Advance Nuclear Technology International, Skultuna, Sweden, 2009

    Google Scholar 

  21. F. Feria and L. Herranz, Creep Assessment of Zry-4 Cladded High Burnup Fuel Under Dry Storage, Prog. Nucl. Energy, 2011, 53(4), p 395–400

    Article  Google Scholar 

  22. P. Bouffioux and N. Rupa, Impact of Hydrogen on Plasticity and Creep of Unirradiated Zircaloy-4 Cladding Tubes, ASTM Spec. Tech. Publ., 2000, 1354, p 399–424

    Google Scholar 

  23. Y.-I. Jung et al., Thermal Creep of Zircaloy-4 Tubes Containing Corrosion-Induced Hydrogen, J. Nucl. Mater., 2011, 419(1), p 213–216

    Article  Google Scholar 

  24. R. Kishore, Effect of Hydrogen on the Creep Behavior of Zr-2.5%Nb Alloy at 723 K, J. Nucl. Mater., 2009, 385(3), p 591–594

    Article  Google Scholar 

  25. D. Setoyama and S. Yamanaka, Indentation Creep Study of Zirconium Hydrogen Solid Solution, J. Alloy. Compd., 2004, 379(1), p 193–197

    Article  Google Scholar 

  26. P. Bouffioux, et al., Interim Dry Storage of PWR Spent Fuel Assemblies Development of a Long Term Creep Law to Assess the Fuel Cladding Integrity, 8th International Conference on Radiactive Waste Management and Environmental Remediation, 2001

  27. Anual Report of Nuclear Power Engineering Corporation, Japan, 2001

  28. A. Sarkar, K. Boopathy, J. Eapen, K. L. Murty, Effect of Hydrogen on Creep Behavior of Zirconium Alloys, LWR Fuel Performance Meeting TopFuel 2013, Charlotte, USA, 2013

  29. Rupa, N., et al. About the Mechanisms Governing the Hydrogen Effect on Viscoplasticity of Unirradiated Fully Annealed Zircaloy-4 Sheet. Thirteenth International Symposium on Zirconium in the Nuclear Industry, 2001

  30. C. Domain, R. Besson, and A. Legris, Atomic-Scale ab Initio Study of the Zr-H System: II. Interaction of H with Plane Defects and Mechanical Properties, Acta Mater., 2004, 52(6), p 1495–1502

    Article  Google Scholar 

  31. V. Mallipudi, S. Valance, and J. Bertsch, Meso-Scale Analysis of the Creep Behavior of Hydrogenated Zircaloy-4, Mech. Mater., 2012, 51, p 15–28

    Article  Google Scholar 

  32. K. Ito, K. Kamimura, Y. Tsukuda, Evaluation of Irradiation Effect on Spent Fuel Cladding Creep Properties. Proceedings of 2004 International Meeting on LWR Fuel Performance, 2004

  33. S. Yamanaka, M. Kuroda, and D. Setoyama, Mechanical Properties of Zirconium Hydride and Hydrogen Solid Solution, Trans. Atom. Energy Soc. Jpn., 2002, 1(4), p 323–327

    Google Scholar 

  34. H.K. Birnbaum and P. Sofronis, Hydrogen-Enhanced Localized Plasticity—A Mechanism for Hydrogen-Related Fracture, Mater. Sci. Eng. A, 1994, 176(1), p 191–202

    Article  Google Scholar 

  35. I. Robertson, The Effect of Hydrogen on Dislocation Dynamics, Eng. Fract. Mech., 2001, 68(6), p 671–692

    Article  Google Scholar 

  36. J. Xu and S.-Q. Shi, Investigation of Mechanical Properties of 〈i〉ε〈/i〉-Zirconium Hydride Using Micro- and Nano-Indentation Techniques, J. Nucl. Mater., 2004, 327(2), p 165–170

    Article  Google Scholar 

  37. M. Puls, S.-Q. Shi, and J. Rabier, Experimental Studies of Mechanical Properties of Solid Zirconium Hydrides, J. Nucl. Mater., 2005, 336(1), p 73–80

    Article  Google Scholar 

  38. K.B. Colas et al., Effect of Thermo-Mechanical Cycling on Zirconium Hydride Reorientation Studied In Situ with Synchrotron x-ray Diffraction, J. Nucl. Mater., 2013, 440(1–3), p 586–595

    Article  Google Scholar 

  39. R.S. Daum, Y.S. Chu, and A.T. Motta, Identification and Quantification of Hydride Phases in Zircaloy-4 Cladding Using Synchrotron x-ray Diffraction, J. Nucl. Mater., 2009, 392(3), p 453–463

    Article  Google Scholar 

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Acknowledgments

Research is funded by DOE/NEUP under the Integrated Research Project (IRP) entitled “Fuel Aging in Storage and Transportation (FAST).”

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Authors and Affiliations

  1. Department of Nuclear Engineering, North Carolina State University, Raleigh, NC, 27695-7909, USA

    A. Sarkar, K. Boopathy, J. Eapen & K. L. Murty

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  1. A. Sarkar
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  2. K. Boopathy
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  3. J. Eapen
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  4. K. L. Murty
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Correspondence to A. Sarkar.

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Sarkar, A., Boopathy, K., Eapen, J. et al. Creep Behavior of Hydrogenated Zirconium Alloys. J. of Materi Eng and Perform 23, 3649–3656 (2014). https://doi.org/10.1007/s11665-014-1129-y

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  • DOI: https://doi.org/10.1007/s11665-014-1129-y

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