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자료유형
학술저널
저자정보
김성우 (한국원자력연구원) 안태영 (한국원자력연구원) 김동진 (한국원자력연구원)
저널정보
한국원자력학회 Nuclear Engineering and Technology Nuclear Engineering and Technology 제53권 제7호
발행연도
2021.7
수록면
2,304 - 2,311 (8page)
DOI
https://doi.org/10.1016/j.net.2021.01.008

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

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