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논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 발행연도
- 2019.1
- 수록면
- 33 - 38 (6page)
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Alloy 600/182 with excellent mechanical/chemical properties have been utilized for nuclear power plants. Although both alloys are known to have superior corrosion resistance, stress corrosion cracking failure hasbeen an issue in primary water environment of nuclear power plants. Therefore, primary water stress corrosioncrack (PWSCC) growth rate tests were conducted to investigate crack growth properties of Alloy 600/182. To investigate PWSCC growth rate, test facilities including water chemistry loop, autoclave, and loadingsystem were constructed. In PWSCC crack growth rate tests, half compact-tension specimens were manufactured. These specimens were then placed inside of the autoclave connected to the loop to provide primary waterenvironment. Tested conditions were set at temperature of 360℃ and pressure of 20MPa. Real time crackgrowth rates of specimens inside the autoclave were measured by Direct Current potential drop (DCPD)method. To confirm inter-granular (IG) crack as a characteristic of PWSCC, fracture surfaces of testedspecimens were observed by SEM. Finally, crack growth rate was derived in a specific stress intensityfactor (K) range and similarity with overseas database was identified.
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참고문헌 (8)
참고문헌 신청
V. N. Shah / 1992 / Nucl. Eng. Design 134:199
S. S. Hwang / 2013 / J. Nucl. Mater 443:321
EPRI / 2018 / Materials Reliability Program, MRP-420, Rev.1
김지현 / 2004 / In-situ Raman Spectroscopic Study of Nickel-base Alloys in Nuclear Power Plants and Its Implications to SCC / CORROSION SCIENCE AND TECHNOLOGY 3(5):198~208
S. S. Kim / 2008 / Corros. and Protec 7:39
S. S. Hwang / 2013 / J. Nucl. Mater 443:321
EPRI / 2018 / Materials Reliability Program, MRP-420, Rev.1
김지현 / 2004 / In-situ Raman Spectroscopic Study of Nickel-base Alloys in Nuclear Power Plants and Its Implications to SCC / CORROSION SCIENCE AND TECHNOLOGY 3(5):198~208
S. S. Kim / 2008 / Corros. and Protec 7:39
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