Skip to main content
SpringerLink
Log in

Materials Assessment for the Canadian SCWR Core Concept

  • Published:
JOM Aims and scope Submit manuscript

Abstract

As part of the development effort of the Generation IV Forum, Canada has undertaken research to support the conceptual design of a pressure-tube-based supercritical water-cooled reactor (SCWR). With an outlet temperature of 625°C and a coolant pressure of 25 MPa, this concept requires fuel-cladding materials that can sustain very harsh in-core conditions. After reviewing the worldwide efforts in SCWR materials since the 1950s, a materials program was created to identify and assess candidate alloys that have potential to operate for 3.5 years as a fuel cladding. An overview of the key aspects and the overall results of this program are presented in this paper, while detailed discussions of individual projects in this program are provided in the accompanying papers of this issue. Further materials R&D work should focus on improving the resistance to stress corrosion cracking, high-temperature strength and ductility as well as microstructural stability.

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

Access this article

Log in via an institution

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

Instant access to the full article PDF.

Institutional subscriptions

Fig. 1
Fig. 2
Fig. 3

Similar content being viewed by others

Notes

  1. For example, see papers in Ref. 35.

References

  1. M. Yetisir, M. Gaudet, and D. Rhodes, Proceedings Of the 6th International Symposium on Supercritical Water-Cooled Reactors (ISSCWR-6) (Shenzhen, China, 3–7 March, 2013), Paper ISSCWR6-13059.

  2. L. Walters, Development of High Performance Pressure Tube Material for Canadian SCWR Concept, JOM. (2016). doi:10.1007/s11837-015-1750-8.

  3. M. Yetisir, R. Xu, M. Gaudet, M. Movassat, H. Hamilton, M. Nimrouzi, and J. Goldak, Proceedings of the 7th International Symposium on Supercritical Water-Cooled Reactors (ISSCWR-7) (Helsinki, Finland, 15–18 March 2015), Paper # ISSCWR7-2072.

  4. D. Guzonas, F. Brosseau, P. Tremaine, J. Meesungnoen, and J.-P. Jay-Gerin, Nucl. Technol. 179, 205 (2012).

    Google Scholar 

  5. D.A. Guzonas, M.K. Edwards, and W. Zheng, Proceedings of the 7th International Symposium on Supercritical Water-Cooled Reactors (ISSCWR-7) (Helsinki, Finland, 15–18 March 2015).

  6. Y. Jiao, W. Zheng, D. Guzonas, and J. Kish, Microstructure Instability of Candidate Fuel Cladding Alloys: Corrosion and Stress Corrosion Cracking Implications, JOM. (2016). doi:10.1007/s11837-015-1739-3.

  7. T.R. Allen, Y. Chen, X. Ren, K. Sridharan, L. Tan, G.S. Was, E. West, D. Guzonas, and G.S. Was, Compr. Nucl. Mater. 5, 279 (2012).

    Article  Google Scholar 

  8. G.P. Gu, W. Zheng, and D. Guzonas, Proceedings of the 2nd Canada-China Joint Workshop on Supercritical Water-Cooled Reactors (CCSC-2010) (Toronto, Canada, 25–28 Apr 2010).

  9. A. Strasser, J. Santucci, K. Lindquist, W. Yario, G. Stern, L. Goldstein, and L. Joseph, An Evaluation of Stainless Steel Cladding for Use in Current Design LWRs, EPRI Report NP 2642, 1982.

  10. I. Emel’yanov, O.A. Shatskaya, E. Yu. Rivkin, and N.Y., Atomnaya Energiya 33(3), 729 (1972) [in Russian] (Translated in Soviet Atomic Energy 33(3), 842 (1972)).

  11. W.E. Ruther, R.R. Schlueter, R.H. Lee, and R.K. Hart, Corrosion 22, 147 (1966).

    Article  Google Scholar 

  12. D.A. Guzonas and W. Cook, Proceedings of the 7th International Symposium on Supercritical Water-Cooled Reactors (ISSCWR-7) (Helsinki, Finland, 15–18 March 2015).

  13. D.A. Guzonas and W.G. Cook, Corros. Sci. 65, 48 (2012).

    Article  Google Scholar 

  14. W. Zheng, Y. Zeng, J. Luo, R. Novotny, J. Li, B. Shalchi Amirkhiz, D. Guzonas, M. Matchim, J. Collier, and L. Yang, The 19th Pacific Basin Nuclear Conference (PBNC 2014) (Hyatt Regency Hotel, Vancouver, Canada, 24–28 August, 2014), Paper PBNC2014-354.

  15. F. Bevilacqua and G.M. Brown, Chloride Deposition from Steam onto Superheater Fuel Clad Materials, General Nuclear Engineering Corporation Report GNEC 295 (1963).

  16. T. Tsuchiya, F. Kano, N. Saito, M. Ookawa, J. Kaneda, and N. Hara, Corrosion 2007 (NACE International, Houston, 2007), Paper No. 07415.

  17. S. Teysseyre and G.S. Was, Corrosion 62, 1101 (2006).

    Article  Google Scholar 

  18. G.S. Was, P. Ampornrat, G. Gupta, S. Teysseyre, E.A. West, T.R. Allen, K. Sridharan, L. Tan, Y. Chen, X. Ren, and C. Pister, J. Nucl. Mater. 371, 176 (2007).

    Article  Google Scholar 

  19. L. Tribouilloy, F. Vaillant, J.-M. Olive, and M. Puiggali, Adv. Mater. Sci. 7, 61 (2007).

    Google Scholar 

  20. S. Lozano-Perez, T. Yamada, T. Terachi, M. Schröder, C.A. English, G.D.W. Smith, C.R.M. Grovenor, and B.L. Eyre, Acta Mater. 57, 5361 (2009).

    Article  Google Scholar 

  21. J. Hou, Q.J. Peng, T. Shoji, J.Q. Wang, E.-H. Han, and W. Ke, Corros. Sci. 53, 1137 (2011).

    Article  Google Scholar 

  22. IAEA, Stress Corrosion Cracking in Light Water Reactors: Good Practices and Lessons Learned. International Atomic Energy Agency Nuclear Energy Series Technical Report NP-T-3.13, Vienna (2011).

  23. Z. Jiao and G.S. Was, J. Nucl. Mater. 408, 246 (2011).

    Article  Google Scholar 

  24. A. Hojná, Corrosion 69, 964 (2013).

    Article  Google Scholar 

  25. P.D. Freyer, T.R. Mager, M.A. Burke, Proceedings of the 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems, (Whistler, Canada, 19–23 April 2007).

  26. F.A. Comprelli, H.J. Busboom, and C.N. Spalaris, Irradiation Effects in Structural Alloys for Thermal and Fast Reactors, ASTM STP 457, (American Society for Testing and Materials, 1969), pp. 400–413.

  27. S.A. Rabin, B.G. Atraz, M.B. Bader, H.J. Busboom, and V.E. Hazel, Examination and Evaluation of Rupture in EVESR Superheat Fuel Rod with 0.012-inch-thick Incoloy-800 Cladding, AEC Research and Development Report GEAP-5416, 1967.

  28. R. Zhou, E.A. West, Z. Jiao, and G.S. Was, J. Nucl. Mater. 395, 11 (2009).

    Article  Google Scholar 

  29. J.S. Janssen, M.M. Morra, and D.J. Lewis, Corrosion 65, 67 (2009).

    Article  Google Scholar 

  30. Y. Jiao, J. Kish, W. Zheng, D. Guzonas, and W. Cook, Proceedings of the 2014 Canada-China Conference on Advanced Reactor Development (CCCARD-2014) (Niagara Falls, Canada, 27–30 April 2014).

  31. S. Xu and B. Shalchi Amirkhiz, Mechanical Property of Fuel Cladding Candidate Alloys for Canadian SCWR Concept, JOM. (2016). doi:10.1007/s11837-015-1725-9.

  32. G.S. Was, Irradiation Creep and Growth, Fundamentals of Radiation Materials Science, 2007, pp. 711–763.

  33. S. Xu, W. Zheng, and L. Yang, CMAT Open Technical Report, 2014-2480-RT, July 2015.

  34. M.L. Grossbeck, K. Ehrlich, and C. Wassilew, J. Nucl. Mater. 174, 264 (1990).

    Article  Google Scholar 

  35. F.A. Garner and N.H. Packan, Radiation-induced Changes in Microstructure: 13th International Symposium, (ASTM International, 1 Jan 1987) and the related ASTM technical publications on Irradiation Effects on Materials.

  36. B.A. Thiele, H. Diehl, W. Ohly, and H. Weber, Nucl. Technol. 66, 597 (1984).

    Google Scholar 

  37. W.G. Johnston, J.H. Rosolowski, A.M. Turkalo, and T. Lauritzen, J. Nucl. Mater. 54, 24 (1974).

    Article  Google Scholar 

  38. F.A. Garner, Chapter 4.02—radiation damage in austenitic steels.Comprehensive Nuclear Materials, ed. R.J.M. Konings (Oxford: Elsevier, 2012), pp. 33–95.

    Chapter  Google Scholar 

  39. F.A. Garner, J. Nucl. Mater. 122, 459 (1984).

    Article  Google Scholar 

  40. M.P. Surh, J.B. Sturgeon, and W.G. Wolfer, J. Nucl. Mater. 328, 107 (2004).

    Article  Google Scholar 

  41. K.P. Boyle, Void swelling assessment of candidate alloys for supercritical water-cooled reactor fuel cladding, CMAT Open Technical Report, 2016-01-RT (2015).

Download references

Author information

Authors and Affiliations

  1. CanmetMATERIALS, Natural Resources Canada, Hamilton, ON, L8P 0A5, Canada

    Wenyue Zheng, Kevin P. Boyle, Jian Li & Su Xu

  2. Canadian Nuclear Laboratories, Chalk River, ON, K0J 1J0, Canada

    David Guzonas

Authors
  1. Wenyue Zheng
    View author publications

    You can also search for this author in PubMed Google Scholar

  2. David Guzonas
    View author publications

    You can also search for this author in PubMed Google Scholar

  3. Kevin P. Boyle
    View author publications

    You can also search for this author in PubMed Google Scholar

  4. Jian Li
    View author publications

    You can also search for this author in PubMed Google Scholar

  5. Su Xu
    View author publications

    You can also search for this author in PubMed Google Scholar

Corresponding author

Correspondence to Wenyue Zheng.

Rights and permissions

Reprints and permissions

About this article

Check for updates. Verify currency and authenticity via CrossMark

Cite this article

Zheng, W., Guzonas, D., Boyle, K.P. et al. Materials Assessment for the Canadian SCWR Core Concept. JOM 68, 456–462 (2016). https://doi.org/10.1007/s11837-015-1758-0

Download citation

  • Received:

  • Accepted:

  • Published:

  • Issue Date:

  • DOI: https://doi.org/10.1007/s11837-015-1758-0

Keywords

Search

Navigation