증기발생기 전열관 틈새복합환경(Pb+S+Cl)에서 Alloy 690의 응력부식균열거동
Stress Corrosion Cracking Behavior of Alloy 690 in Crevice Environment (Pb + S + Cl) in a Steam Generator Tube
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초록
The secondary coolant of a nuclear power plant has small amounts of various impurities (S, Pb, and Cl,etc.) introduced during the initial construction, maintenance, and normal operation. While the concentrationof impurities in the feed water is very low, the flow of the cooling water is restricted, so impurities canaccumulate on the Top of Tubesheet (TTS). This environment is chemically very complicated and has avery wide range of pH from acidic to alkaline. In this study, the characteristics of the oxide and the mechanismof stress corrosion cracking (SCC) are investigated for Alloy 690 TT in alkaline solution containing Pb,Cl, and S. Reverse U-bend (RUB) specimens were used to evaluate the SCC resistance. The test solutioncomprises 3m NaCl + 500ppm Pb + 0.31m Na2SO4 + 0.45m NaOH. Experimental results show that Alloy690 TT of the crevice environment containing Pb, S, and Cl has significant cracks, indicating that Alloy690 is vulnerable to stress corrosion cracking under this environment.
Abstract
The secondary coolant of a nuclear power plant has small amounts of various impurities (S, Pb, and Cl,etc.) introduced during the initial construction, maintenance, and normal operation. While the concentrationof impurities in the feed water is very low, the flow of the cooling water is restricted, so impurities canaccumulate on the Top of Tubesheet (TTS). This environment is chemically very complicated and has avery wide range of pH from acidic to alkaline. In this study, the characteristics of the oxide and the mechanismof stress corrosion cracking (SCC) are investigated for Alloy 690 TT in alkaline solution containing Pb,Cl, and S. Reverse U-bend (RUB) specimens were used to evaluate the SCC resistance. The test solutioncomprises 3m NaCl + 500ppm Pb + 0.31m Na2SO4 + 0.45m NaOH. Experimental results show that Alloy690 TT of the crevice environment containing Pb, S, and Cl has significant cracks, indicating that Alloy690 is vulnerable to stress corrosion cracking under this environment.
참고문헌 (13)
- D. Gomez-Briceno, , Nucl. Eng. Des., 165, pp.161~ (1996)[DOI]
- R. S. Dutta, , J. Nucl. Mater., 393, pp.343~ (2009)[DOI]
- R. L. Tapping, , Proc.5th CNS International Steam Generator Conf. (2006)
- R. W. Staehle, , Corrosion, 59, pp.931~ (2003)[DOI]
- R. W. Staehle, , Corrosion, 60, pp.5~ (2004)[DOI]
- R. W. Staehle, , Corrosion, 60, pp.115~ (2004)[DOI]
- S. S. Hwang, , J. Corros. Sci. Soc. of Kor., 25, pp.327~ (1996)
- T. Sakai, , Proc. 10th International Sympsium on Environmental Degradation of Materials in Nuclear Power System-Water Reactors, pp.12~37 (1989)
- Z. J. Bai, , Corros. Sci. Tech., 31, pp.34~ (2002)
- M. G. Burke, , Proc. 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, pp.851~873 (2007)
- T. Y Ahn, , Proc. 18th International Conf. on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, pp.501~507 (2017)
- 김동진, "Susceptibility of Alloy 690 to Stress Corrosion Cracking in Caustic Aqueous Solutions", Nuclear Engineering and Technology, 45(1), pp.67~72 (2013), 한국원자력학회
- 김동진, "고온, 고압 알칼리 수용액에서의 Alloy 600 산화막 특성에 미치는 납 농도 영향", CORROSION SCIENCE AND TECHNOLOGY, 11(3), pp.96~102 (2012), 한국부식방식학회[DOI]