Corrosion Science and Technology

The Corrosion Science Society of Korea (한국부식방식학회)
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Stress Corrosion Cracking Behavior of Alloy 690 in Crevice Environment (Pb + S + Cl) in a Steam Generator Tube

증기발생기 전열관 틈새복합환경(Pb+S+Cl)에서 Alloy 690의 응력부식균열거동

  • Shin, Jung-Ho (Nuclear Materials Research Division, Korea Atomic Energy Research Institute (KAERI)) ;
  • Lim, Sang-Yeop (Nuclear Materials Research Division, Korea Atomic Energy Research Institute (KAERI)) ;
  • Kim, Dong-Jin (Nuclear Materials Research Division, Korea Atomic Energy Research Institute (KAERI))
  • 신정호 (한국원자력연구원 원자력재료연구부) ;
  • 임상엽 (한국원자력연구원 원자력재료연구부) ;
  • 김동진 (한국원자력연구원 원자력재료연구부)
  • Received : 2014.10.30
  • Accepted : 2015.02.23
  • Published : 2018.06.29
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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 concentration of impurities in the feed water is very low, the flow of the cooling water is restricted, so impurities can accumulate on the Top of Tubesheet (TTS). This environment is chemically very complicated and has a very wide range of pH from acidic to alkaline. In this study, the characteristics of the oxide and the mechanism of 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 solution comprises 3m NaCl + 500ppm Pb + 0.31m $Na_2SO_4$ + 0.45m NaOH. Experimental results show that Alloy 690 TT of the crevice environment containing Pb, S, and Cl has significant cracks, indicating that Alloy 690 is vulnerable to stress corrosion cracking under this environment.

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References

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