A primary water stress corrosion cracking (SCC) and an outside diameter SCC have occurred in the steam generator (SG) tubes of nuclear power plants around the world. It is important to establish the repair criteria for the degraded tubes to assure a reactor integrity, and yet maintain the plugging ratio within the limits needed for an efficient operation. For assessment and management of the degradation, it became crucial to understand initial leak behaviors under a small pressure and leak rate evolution under a constant pressure of SCC flaws. Stress corrosion cracked tube specimens were prepared by using a room temperature cracking technique, and leak behaviors of these tubes were measured at room temperature. Water pressure inside the tube was increased slowly in a step like manner with a designated holding time. Water leak rates just after a ligament rupture were measured by collecting the leaked water in a plastic container for a designated time. A leak rate was calculated by dividing the amount of water by the time. Under 3.45 MPa, a small water droplet was formed, but it did not grow after a 10 minute holding period at a constant pressure of 3.45 MPa. A throughwall crack seemed to open at around 8.28 MPa (1200 psi). Some tubes with 100 % through wall cracks did not show a leakage at 10.8 MPa, which is a typical pressure difference of pressurized water reactors (PWRs) during a normal operation. The higher the pressure was applied, the larger the rates of increase with the time were. Axial cracks showed a lower leak pressure than that of the circumferential cracks, which might be from a higher hoop stress than the axial stress. A large open and long axial crack showed an increasing leak rate with the time at a constant pressure.