누설 및 파열실험용 SCC 결함 전열관 제작 및 누설거동 평가
Production of SCC Flaws and Evaluation Leak Behavior of Steam Generator Tubes
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초록
A forced outage due to a steam generator tube leak in a Korean nuclear power plant was reported.1) Primary water stress corrosion cracking has occurred in many tubes in the plant, and they were repaired using sleeves or plugs. In order to develop proper repair criteria, it is necessary to understand the leak behavior of the tubes containing stress corrosion cracks. Stress corrosion cracks were developed in 0.1 M sodium tetrathionate solution at room temperature. Steam generator(SG) tubes with short cracks were successfully fabricated with a restricted solution contact method. The leak rates of the degraded tubes were measured at room temperature. Some tubes with 100 % through wall cracks showed an increase of leak rate with time at a constant pressure.
Abstract
A forced outage due to a steam generator tube leak in a Korean nuclear power plant was reported.1) Primary water stress corrosion cracking has occurred in many tubes in the plant, and they were repaired using sleeves or plugs. In order to develop proper repair criteria, it is necessary to understand the leak behavior of the tubes containing stress corrosion cracks. Stress corrosion cracks were developed in 0.1 M sodium tetrathionate solution at room temperature. Steam generator(SG) tubes with short cracks were successfully fabricated with a restricted solution contact method. The leak rates of the degraded tubes were measured at room temperature. Some tubes with 100 % through wall cracks showed an increase of leak rate with time at a constant pressure.
참고문헌 (7)
- Destructive Analysis on Pulled Tubes from Ulchin Unit 4-Final Report, "J.S. Kim", Atomic Energy Research Institute, ~ (2003)
- Seong Sik Hwang, "SCC Mechanism of Ni Base alloys in Lead Contaminated Water", CORROSION SCIENCE AND TECHNOLOGY, 7(3), pp.189~193 (2008), 한국부식방식학회
- JoungSooKim, "Effect of the Microstructure of Inconel 600 HTMA on Susceptibility to Stress Corrosion Cracking on the Simulated Nuclear Power Plant Operating Conditions", CORROSION SCIENCE AND TECHNOLOGY, 31(4), pp.320~326 (2002), 한국부식방식학회
- HongPyoKim, "Stress Corrosion Cracking of Alloy 600 and Alloy 690 in Caustic Solution", CORROSION SCIENCE AND TECHNOLOGY, 32(2), pp.2~87 (2003), 한국부식방식학회
- S.S.Hwang, , Corrosion and Protection, 7, pp.7~ (2008)
- V.N. Shah, "Aging and life extension of major light water reactor components" (1993), Elsevier
- Second U.S.Nuclear Regulatory Commission International Steam Generator Tube Integrity Research Program-Final Project Report, "S. Bakhtiari", 90~, pp.NUREG CR-6804 (2003), ANL-02[DOI]