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DOI QR Code

SUSCEPTIBILITY OF ALLOY 690 TO STRESS CORROSION CRACKING IN CAUSTIC AQUEOUS SOLUTIONS

  • Kim, Dong-Jin (Nuclear Materials Division, Korea Atomic Energy Research Institute (KAERI)) ;
  • Kim, Hong Pyo (Nuclear Materials Division, Korea Atomic Energy Research Institute (KAERI)) ;
  • Hwang, Seong Sik (Nuclear Materials Division, Korea Atomic Energy Research Institute (KAERI))
  • Received : 2012.03.23
  • Accepted : 2012.05.26
  • Published : 2013.02.25

Abstract

Stress corrosion cracking (SCC) behaviors of Alloy 690 were studied in lead-containing aqueous alkaline solutions using the slow strain rate tension (SSRT) tests in 0.1M and 2.5M NaOH with and without PbO at $315^{\circ}C$. The side and fracture surfaces of the alloy were then examined using scanning electron microscopy after the SSRT test. Microstructure and composition of the surface oxide layer were analyzed by using a field emission transmission electron microscopy, equipped with an energy dispersive X-ray spectroscopy. Even though Alloy 690 was almost immune to SCC in 0.1M NaOH solution, irrespective of PbO addition, the SCC resistance of Alloy 690 decreased in a 2.5M NaOH solution and further decreased by the addition of PbO. Based on thermodynamic stability and solubility of oxide, high Cr of 30wt% in the Alloy 690 is favorable to SCC in mild alkaline and acidic solutions whereas the SCC resistance of high Cr Alloy 690 is weakened drastically in the strong alkaline solution where the oxide is not stable any longer and solubility is too high to form a passive oxide locally.

Keywords

References

  1. J. Benson, 29th Annual EPRI Steam Generator NDE Workshop, July 12-14, 2010.
  2. NUREG report, NUREG/CP-0189, 2003.
  3. A. Baum, K. Evans, Proc. of 12th Int. Symp. on Environmental Degradation of Materials in Nuclear Power Systems- Water Reactors, Salt Lake City, Utah, Aug. 1-5, 2005, p. 1155-1162.
  4. J. M. Sarver, EPRI Workshop on Intergranular Corrosion and Primary Water Stress Corrosion Cracking Mechanisms, NP-5971, EPRI, Palo Alto, 1987, p. C11/1.
  5. M. L. Castano-Marin, D. Gomez-Briceno, F. Hernandez- Arroyo, Proc. of 6th Int. Symp. on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, San Diego, CA, Aug. 1-5, 1993, p. 189-196.
  6. M. D. Wright, M. Mirzai, Proc. of 9th Int. Symp. on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, Newport Beach, CA, Aug. 1-5, 1999, p. 657-665.
  7. R. W. Staehle, Proc. of 11th Int. Symp. on Environmental Degradation of Materials in Nuclear Power Systems- Water Reactors, Stevenson, WA, Aug. 10-14, 2003, p. 381-422.
  8. K. Fruzzetti, Workshop of Effects of Pb and S on the Performance of Secondary Side Tubing of Steam Generators in PWRs, ANL, IL, May 24-27, 2005.
  9. D.-J. Kim, H. C. Kwon, H. W. Kim, S. S. Hwang, H. P. Kim, Corrosion Science, 53 (2011) 1247-1253. https://doi.org/10.1016/j.corsci.2010.12.016
  10. D.-J. Kim, H. W. Kim, S. W. Kim, H. P. Kim, Rev. Adv. Mater. Sci., 28 (2011) 64-68.
  11. F. Vaillant, D. Buisine, B. Prieux, D. Gomez Briceno, L. Castano, Eurocorr 96, Nice, 1996, p. 13/1.
  12. U. C. Kim, K. M. Kim, E. H. Lee, J. Nucl. Mater., 341 (2005) 169-174. https://doi.org/10.1016/j.jnucmat.2005.01.018
  13. Y. Yi, S. Eom, H. Kim, J. Kim, J. Nucl. Mater., 347 (2005) 151-160. https://doi.org/10.1016/j.jnucmat.2005.08.011
  14. D.-J. Kim, H. P. Kim, S. S. Hwang, J. S. Kim, J. Park, Met. Mater. Int., 16 (2010) 259-266. https://doi.org/10.1007/s12540-010-0415-y
  15. D.-J. Kim, Y. S. Lim, H. C. Kwon, S. S. Hwang, H. P. Kim, J. Nanoscience and Nanotechnology, 10 (2010) 85-91. https://doi.org/10.1166/jnn.2010.1530
  16. HSC Chemistry Database, 6.0.

Cited by

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  2. Effects of Lead on the Initial Corrosion Behavior of 316LN Stainless Steel in High-Temperature Alkaline Solution pp.2194-1289, 2018, https://doi.org/10.1007/s40195-018-0821-6
  3. Study on electrochemical behavior of 690 alloy with corrosion products in simulated PWR primary water environment pp.0003-5599, 2018, https://doi.org/10.1108/ACMM-07-2018-1961
  4. Nanoscale Precursor Sites and their Importance in the Prediction of Stress Corrosion Cracking Failure vol.75, pp.3, 2019, https://doi.org/10.5006/2928