Affiliation:
1. Korea Atomic Energy Research Institute
Abstract
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.
Publisher
Trans Tech Publications, Ltd.
Subject
Mechanical Engineering,Mechanics of Materials,General Materials Science
Cited by
2 articles.
订阅此论文施引文献
订阅此论文施引文献,注册后可以免费订阅5篇论文的施引文献,订阅后可以查看论文全部施引文献