Nuclear Engineering and Technology

  • 제54권12호
  • /
  • Pages.4481-4490
  • /
  • 2022
  • /
  • 1738-5733(pISSN)
  • /
  • 2234-358X(eISSN)
한국원자력학회 (Korean Nuclear Society)
권호 표지
DOI QR코드

DOI QR Code

New test method for real-time measurement of SCC initiation of thin disk specimen in high-temperature primary water environment

  • Geon Woo Jeon (Material Safety Technology Development Division, Korea Atomic Energy Research Institute) ;
  • Sung Woo Kim (Material Safety Technology Development Division, Korea Atomic Energy Research Institute) ;
  • Dong Jin Kim (Material Safety Technology Development Division, Korea Atomic Energy Research Institute) ;
  • Chang Yeol Jeong (Department of Nuclear and Energy System Engineering, Dongguk University)
  • 투고 : 2022.04.27
  • 심사 : 2022.07.25
  • 발행 : 2022.12.25
PDF 다운로드
⟨ 이전 논문 다음 논문 ⟩

초록

In this study, a new rupture disk corrosion test (RDCT) method was developed for real-time detection of stress corrosion cracking (SCC) initiation of Alloy 600 in a primary water environment of pressurized water reactors. In the RDCT method, one side of a disk specimen was exposed to a simulated primary water at high temperature and pressure while the other side was maintained at ambient pressure, inducing a dome-shaped deformation and tensile stress on the specimen. When SCC occurs in the primary water environment, it leads to the specimen rupture or water leakage through the specimen, which can be detected in real-time using a pressure gauge. The tensile stress applied to the disk specimen was calculated using a finite element analysis. The tensile stress was calculated to increase as the specimen thickness decreased. The SCC initiation time of the specimen was evaluated by the RDCT method, from which result it was found that the crack initiation time decreased with the decrease of specimen thickness owing to the increase of applied stress. After the SCC initiation test, many cracks were observed on the specimen surface in an intergranular fracture mode, which is a typical characteristic of SCC in the primary water environment.

키워드

과제정보

This research was supported by the Korean Nuclear R&D Program 2021M2E4A1037979) through the National Research Foundation of Korea (NRF) funded by the Ministry of Science, and partially funded by the Korean R&D program (RS-2022-00143718). The authors thank Mr. Tae-Young Kim for complementary FE analysis.

참고문헌

  1. D. Feron, J.M. Olive (Eds.), Corrosion Issues in Light Water Reactors - Stress Corrosion Cracking, Woodhead Publishing Ltd., New York, 2007.
  2. M. Le Calvar, I. De Curieres, Corrosion issues in pressurized water reactor (PWR) systems, Nucl, Corros. Sci. Eng (2012) 473-547.
  3. R.W. Staehle, Quantitative micro-nano (QMN) approach to SCC mechanism and prediction-starting a third meeting, in: Proc. 15th Int. Conf. On Env. Deg. Mater. Nucl. Power Syst, - Water Reactors, Colorado Springs, 2011, pp. 1535-1625.
  4. Y.S. Garud, A Simplified Model for Assessment of SCC Initiation Time in Alloy 600, EPRI, TR-109137, 1997. Palo Alto.
  5. Y.S. Garud, Validation of Stress Corrosion Cracking Initiation Model for Stainless Steel and Nickel Alloys: Effects of Cold Work, EPRI, TR-1025121, Palo Alto, 2012.
  6. Z. Zhai, M.B. Toloczko, M.J. Olszta, S.M. Bruemmer, Stress corrosion crack initiation of alloy 600 in PWR primary water, Corrosion Sci. 123 (2017) 76-87. https://doi.org/10.1016/j.corsci.2017.04.013
  7. D.A. Jones, Principles and Prevention of Corrosion (2nd Ed.), Prentice Hall, Upper Saddle River, 1995.
  8. ASTM International, Standard Practice for Making and Using U-Bend Stress-Corrosion Test Specimens, 2003. ASTM G30-97, West Conshohocken.
  9. JSA, Stress Corrosion Cracking Testing of Metals and Alloy Using Reverse U-Bend Test Method, JIS G 0511, 2006. Tokyo.
  10. ASTM International, Standard Practice for Making and Using C-Ring Stress-Corrosion Test Specimens, 2007. ASTM G38-01, West Conshohocken.
  11. P.L. Andresen, L.M. Young, G.M. Catlin, P.W. Emich, G.M. Gordon, Stress-corrosion-crack initiation and growth-rate studies on titanium grade 7 and Alloy 22 in concentrated groundwater, Metall. Mater. Trans. 36A (2005) 1187-1198.
  12. E. Richey, D.S. Morton, M.K. Schurman, SCC initiation testing of nickel-based alloys using in-situ monitored uniaxial tensile specimens, in: Proc. 12th Int. Conf. On Env. Deg. Mater. Nucl. Power Syst, Water Reactors, Salt Lake City, 2005, pp. 947-956.
  13. K. Takakura, K. Nakata, K. Fujiomoto, K. Sakima, N. Kubo, IASCC properties of cold worked 316 stainless steel in PWR primary water, in: Proc. 14th Int. Conf. On Env. Deg. Mater. Nucl. Power Syst, Water Reactors, Virginia Beach, 2009, pp. 1207-1218.
  14. S. Pemberton, J. Beswick, M. Chatterton, J. Meadows, S. Medway, Proc.. in: Proc. 19th Int. Conf. On Env. Deg. Mater. Nucl. Power Syst, Water Reactors, Boston, 2019, pp. 346-355.
  15. P.J. Meadows, P.L. Andresen, M.B. Toloczko, W.-J. Kuang, S. Ritter, M. Bjurman, L. Zhang, M. Ernestova, A. Toivonen, F. Perosanz-Lopez, J.W. Stairmand, K.J. Mottershead, International round-robin on stress corrosion crack initiation of Alloy 600 material in pressurized water reactor primary water, Corrosion 76 (2020) 719-733. https://doi.org/10.5006/3532
  16. ASTM International, Standard Test Method for Measurement of Fatigue Crack Growth Rates, 2005. ASTM E 647-05, West Conshohocken.
  17. P.L. Andresen, M.M. Morra, K. Ahluwalia, J. Wilson, Proc.. in: Proc. 14th Int. Conf. On Env. Deg. Mater. Nucl. Power Syst, Water Reactors, Virginia, 2009, pp. 846-887.
  18. D.J. Kim, H.P. Kim, S.S. Hwang, Susceptibility of alloy 690 to stress corrosion cracking in caustic aqueous solutions, Nucl. Eng. Technol. 45 (2013) 67-72. https://doi.org/10.5516/NET.07.2012.021
  19. S.M. Bruemmer, M.J. Olszta, N.R. Overman, M.B. Toloczko, Cold-work effects on stress corrosion cracking growth in Alloy 690 tubing and plate materials, in: Proc. 17th Int. Conf. On Env. Deg. Mater. Nucl. Power Syst, - Water Reactors, Ontario, 2015, pp. 1-17.
  20. Y.S. Lim, D.J. Kim, S.W. Kim, H.P. Kim, Crack growth and cracking behavior of Alloy 600/182 and Alloy 690/152 welds in simulated PWR primary water, Nucl. Eng. Technol. 51 (2019) 228-237. https://doi.org/10.1016/j.net.2018.09.011
  21. S.W. Kim, K.H. Eom, Y.S. Lim, D.J. Kim, PWSCC growth rate model of alloy 690 for head penetration nozzles of Korean PWRs, Nucl. Eng. Technol. 51 (2019) 1060-1068. https://doi.org/10.1016/j.net.2019.01.010
  22. T. Moss, W. Kuang, G.S. Was, Stress corrosion crack initiation in Alloy 690 in high temperature water, Curr. Opin. Solid State Mater. Sci. 22 (2018) 16-25. https://doi.org/10.1016/j.cossms.2018.02.001
  23. Chikezie Nwaoha, Process Plant Equipment: Operation, Control, and Reliability, first ed., John Wiley & Sons, Inc, 2012.
  24. P. Scott, P. Combrade, R. Kilian, A. Roth, P. Andresen, Y. Kim, Status review of initiation of environmentally assisted cracking and short crack growth, EPRI report (2005), 1011788. Palo Alto.
  25. V.I. Vodyanik, Design of safety rupture discs, Chem. Petrol. Eng. 8 (1972) 586-588. https://doi.org/10.1007/BF01139117
  26. ASTM Interntional, Standard Test Methods for Determining Average Grain Size, 2010. ASTM E112-10, West Conshohocken.
  27. H. Zhu, W. Xu, Z. Luo, H. Zheng, Finite element analysis on the temperature-dependent burst behavior of domed 316l austenitic stainless steel rupture disc, Metals 10 (2020) 1-11.
  28. R.W. Werne, Stress Analysis of a Rupture Disk, UCID-16761, University of California, 1975.
  29. I. Simonovski, S. Holmstroem, M. Bruchhausen, Small punch tensile testing of curved specimens: finite element analysis and experiment, Int. J. Mech. Sci. 120 (2017) 204-213. https://doi.org/10.1016/j.ijmecsci.2016.11.029
  30. X. Kong, J. Zhang, X. Li, Z. Jin, H. Zhong, Y. Zhan, F. Han, Procedia Manuf. 15 (2018) 892-898. https://doi.org/10.1016/j.promfg.2018.07.408
  31. J. Peng, V.D. Vijayannd, D. Knowles, C. Truman, M. Mostafavi, Int. J. Pres. Ves. Pip. 193 (2021), 104468.