Nuclear Engineering and Technology

  • 제53권7호
  • /
  • Pages.2304-2311
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  • 2021
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  • 1738-5733(pISSN)
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  • 2234-358X(eISSN)
한국원자력학회 (Korean Nuclear Society)
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Analysis of dislocation density in strain-hardened alloy 690 using scanning transmission electron microscopy and its effect on the PWSCC growth behavior

  • Kim, Sung-Woo (Materials Safety Research Division, Korea Atomic Energy Research Institute) ;
  • Ahn, Tae-Young (Materials Safety Research Division, Korea Atomic Energy Research Institute) ;
  • Kim, Dong-Jin (Materials Safety Research Division, Korea Atomic Energy Research Institute)
  • 투고 : 2020.10.27
  • 심사 : 2021.01.07
  • 발행 : 2021.07.25
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초록

The dislocation density in strain-hardened Alloy 690 was analyzed using scanning transmission electron microscopy (STEM) to study the relationship between the local plastic strain and susceptibility to primary water stress corrosion cracking (PWSCC) in nuclear power plants. The test material was cold-rolled at various thickness reduction ratios from 10% to 40% to simulate the strain-hardening condition of plant components. The dislocation densities were measured at grain boundaries (GB) and in grain interiors of strain-hardened specimens from STEM images. The dislocation density in the grain interior monotonically increased as the strain-hardening proceeded, while the dislocation density at the GB increased with strain-hardening up to 20% but slightly decreases upon further deformation to 40%. The decreased dislocation density at the GB was attributed to the formation of deformation twins. After the PWSCC growth test of strain-hardened Alloy 690, the fraction of intergranular (IG) fracture was obtained from fractography. In contrast to the change in the dislocation density with strain-hardening, the fraction of IG fracture increased remarkably when strain-hardened over 20%. From the results, it was suggested that the PWSCC growth behavior of strain-hardened Alloy 690 not only depends on the dislocation density, but also on the microstructural defects at the GB.

키워드

과제정보

This research was supported by the Korean Nuclear R&D Program through the National Research Foundation of Korea (NRF) funded by the Ministry of Science and ICT (2017M2A8A4015155).

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