Top > JDR > Generation IV Material Issues - Case SCWR

single-dr.php

JDR Vol.5 No.4 pp. 469-478
(2010)
doi: 10.20965/jdr.2010.p0469

Paper:

Views over last 60 days: 525

Generation IV Material Issues - Case SCWR

Sami Penttilä*, Aki Toivonen*, Laura Rissanen*,
and Liisa Heikinheimo**

*Materials and Building, Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT, Espoo, Finland

**Teollisuuden Voima Oyj (TVO), Töölönkatu 4, 00100 Helsinki, Finland

Received:
March 24, 2010
Accepted:
July 2, 2010
Published:
August 1, 2010
Creative Commons License
Keywords:
GenIV, supercritical water, corrosion, stress corrosion cracking, creep, austenitic alloy, ODS steel
Abstract
Six Generation IV nuclear power concepts becoming global research topics accepted by the GenIV international forum (GIF) with common objectives of promoting technological efficiency and safety provide attractive features and demanding new challenges e.g., increased operating temperatures, higher irradiation doses, more aggressive coolants, and/or longer life expectations than GenII and GenIII plants. This paper reviews the performance of commercial candidate materials for in-core super-critical water reactor (SCWR) applications focusing on corrosion, stress corrosion cracking (SCC), and creep issues based on work within the Euratom phase 2 high performance light water reactor phase 2 (HPLWR) project. General corrosion, i.e., oxidation rate tests, and SCC tests have been done on selected iron- and nickel-based alloys at 500°C and 650°C in supercritical water (SCW) at a pressure of 25 MPa with an oxygen concentration of 125-150 ppb in all tests. Constant load creep tests have been done on selected austenitic stainless steel at 650°C in SCW at 25 MPa and 1 ppm O2 and in inert atmospheres of He and 0.1 MPa. Based on materials studies done during the HPLWR Phase2 project, current candidates for European SCWR core internals are austenitic stainless steels having sufficient oxidation and creep resistance up to 500°C or 550°C. High chromium austenitic steels and ferritic/martensitic ODS steels are considered, in longterm, for fuel rod cladding thanks to their good oxidation resistances up to 650°C.
Cite this article as:
S. Penttilä, A. Toivonen, L. Rissanen, and L. Heikinheimo, “Generation IV Material Issues - Case SCWR,” J. Disaster Res., Vol.5 No.4, pp. 469-478, 2010.
Data files:
References
  1. [1] World Nuclear Association, “The Nuclear Renaissance,” 2009,
    http://www.world-nuclear.org/info/inf104.html
  2. [2] U.S. DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, “A Technology Roadmap for Generation IV Nuclear Energy Systems,” GIF-002-00, 03-GA50034, 2002.
    http://www.gen-4.org/PDFs/GenIVRoadmap.pdf
  3. [3] K. Fischer, T. Schulenberg, and E. Laurien, “Design of a supercritical water-cooled reactor with a three-pass core arrangement,” Nuclear Engineering and Design, Vol.239, pp. 800-812, 2009.
  4. [4] L. Zhang, F. Zhu, and R. Tang, “Corrosion mechanisms of candidate structural materials for supercritical water-cooled reactor,” Front. Energy Power Eng., Vol.3, pp. 233-240, 2009.
  5. [5] K. Sridharan, A. Zillmer, J. R. Licht, T. R. Allen, M. H. Anderson, and L. Tan, “Corrosion behavior of candidate alloys for supercritical water reactors,” ICAPP 04, Pittsburgh, PA, p. 537, Paper 4136. 2004.
  6. [6] Y. H. Jeong, J. Y. Park, H. G. Kim et al., “Corrosion of zirconiumbased fuel cladding alloys in supercritical water,” 12th Int. Conf. on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, Salt Lake City, USA. pp. 1369-1378, 2006.
  7. [7] C. Sun et al., “Progress in corrosion resistant materials for supercritical water reactors,” Corrosion Science, Vol.51, pp. 2508-2523, 2009.
  8. [8] D. M. Bartels, M. Anderson, P. Wilson, T. Allen, and K. Sridharan, “Supercritical water radiolysis chemistry,” Supercritical water corrosion,
    Available from: http://nuclear.inel.gov/deliverables/docs/uwnd_scw_level_ii_sep_2006_v3.pdf .
  9. [9] 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, “Corrosion and stress corrosion cracking in supercritical water,” J. of Nuclear Materials, Vol.371, pp. 176-201, 2007.
  10. [10] S. Penttilä, A. Toivonen, L. Heikinheimo, and R. Novotny, “Corrosion studies of candidate materials for European HPLWR,” Proc. of ICAPP ’08, Anaheim, CA, USA, June 8-12, Paper 8164, 2008.
  11. [11] S. Penttilä, A. Toivonen, L. Heikinheimo, and R. Novotny, “SCC properties and oxidation behaviour of austenitic alloys at supercritical water conditions,” 4th Int. Symposium on Supercritical Water-Cooled Reactors, 8-11 March, Heidelberg, Germany, Paper No.60, 2009.
  12. [12] P. J. Maziasz, J. P, Shingledecker, P. A. Pint, N. D. Evans, Y. Yamamoto, K.More, and E. Lara-Curzio, “Overview of creep strength and oxidation of heat-resistant alloy sheets and foils for compact heat exchangers,” Proc. of GT2005, ASME Turbo Expo 2005: Power for Land, Sea and air, June 6-9, reno-Tahoe, Nevada, USA, GT2005-68927, 2005.
  13. [13] I. Betova, M. Bojinov, P. Kinnunen, V. Lehtovuori, S. Peltonen, S. Penttilä, and T. Saario, “Composition, Structure, and Properties of Corrosion Layers on Ferritic and Austenitic Steels in Ultrasupercritical Water,” J. of The Electrochemical Society, 153 (11) B464-B473, 2006.
  14. [14] G. R. Holcomb and D. E. Alman, “The effect of manganese additions on the reactive evaporation of chromium in Ni-Cr alloys,” Scripta Materialia, Vol.54, pp. 1821-1825, 2006.
  15. [15] R. Viswanathan, J. Sarver, and J. M. Tanzosh, “Boiler materials for Ultra-Supercritical Coal Power plants-Steamside oxidation,” J. of Materials Engineering and Performance, Vol.15, pp. 255-273, 2006.
  16. [16] L. Tan, K. Sridharan, T. Allen, R. Nanstad, and D. McClintock, “Microstructure tailoring for property improvements by grain boundary engineering,” J. of Nuclear Materials, Vol.374, pp. 270-280, 2008.
  17. [17] L. Tan, X. Ren, K. Sridharan, and T. Allen, “Effect of shot-peening on the oxidation of alloy 800H exposed to supercritical water and cyclic oxidation,” Corrosion Science, Vol.50, pp. 2040-2046, 2008.
  18. [18] W. Zheng, D. Guzonas, J. Wills, and J. Li, “Novel approach to the development of in-core materials for a supercritical water cooled reactor,” 16th Pacific Basin Nuclear Conf. (16PNC), 13-18 October, Aomori, Japan. Paper ID P16P1413, 2008.
  19. [19] S. Teysseyre, and G. Was, “Stress Corrosion Cracking of Austenitic Alloys in Supercritical Water,” Corrosion, Vol.62, No.12, 2006.
  20. [20] T. Lau, X. Lin, S. Yip, and K. Van Vliet, “Atomistic examination of the unit processes and vacancy-dislocation interaction in dislocation climb,” Scripta Materialia, Vol.60, pp. 399-402, 2009.
  21. [21] E. Andrieu, B. Pieraggi, and A. F. Gourgues, “Role of metal-oxide interfacial reactions on the interactions between oxidation and deformation,” Scripta Materialia, Vol.39, Nos.4/5, pp. 597-601, 1998.
  22. [22] www.matweb.com, Plansee PM2000 technical data sheet.

Creative Commons License  This article is published under a Creative Commons Attribution-NoDerivatives 4.0 Internationa License.

*This site is desgined based on HTML5 and CSS3 for modern browsers, e.g. Chrome, Firefox, Safari, Edge, Opera.

Last updated on Apr. 18, 2024