Primary water stress corrosion cracking (SCC) and an outside diameter SCC have occurred in the steam generator (SG) tubes of nuclear power plants in the republic of Korea and around the world. Although high corrosion resistant alloy 690 has been replacing the alloy 600 tubings, it is important to establish the repair criteria for the remaining degraded alloy 600 tubings to reassure regarding the reactors integrity, and still maintain the plugging ratio within the limits needed for its efficient operations. For assessment and management of the degradation, it is crucial to understand the initial leak behaviors and time dependent leak rate change from SCC flaws under a constant pressure. Stress corrosion cracked tube specimens were prepared at room temperature by the contact of sodium tetrathionate solution. The initial leak rate and time dependent leak rate were measured at different pressure ranges with time. Water pressure inside the tube was slowly increased in a step like manner with a designated holding time. The leak rate was calculated by dividing the amount of water by the time. A large open and long axial crack showed an increasing leak rate with time at a constant pressure, whereas small opened cracks did not show an increase in a time dependent leak rate. Under some pressures, the leak rate did not increase with the increase of pressure due to a tightness of circumferential cracks. Throughwall axial crack of 5 mm long may exhibit the leakage of action level 1 of the EPRI leakage guideline.

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