Rupture pressure of wear degraded alloy 600 steam generator tubings

• Published in: Journal of nuclear materials 2008-02, Vol.373 (1), p.71-74
• Main Authors: Hwang, Seong Sik, Namgung, Chan, Jung, Man Kyo, Kim, Hong Pyo, Kim, Joung Soo
• Format: Article
• Language: English
• Subjects: Applied sciences; Controled nuclear fusion plants; Energy; Energy. Thermal use of fuels; Exact sciences and technology; Fission nuclear power plants; Fuels; Installations for energy generation and conversion: thermal and electrical energy; Nuclear fuels
• ISSN: 0022-3115; 1873-4820
• DOI: 10.1016/j.jnucmat.2007.05.020

Fretting/wear degradation at the tube support in the U-bend region of a steam generator (SG) of a pressurized water reactor (PWR) has been reported. Simulated fretted flaws were machined on SG tubes of 195 mm in length. A pressure test was carried out with the tubes at room temperature by using a high pressure test facility which consisted of a water pressurizing pump, a test specimen section and a control unit. Water leak rates just after a ligament rupture or a burst were measured. Tubes degraded by up to 70% of the tube wall thickness (TW) showed a high safety margin in terms of the burst pressure during normal operating conditions. Tubes degraded by up to 50% of the TW did not show burst. Burst pressure depended on the defect depths rather than on the wrap angles. The tube with a wrap angle of 0° showed a fish mouth fracture, whereas the tube with a 45° wrap angle showed a three way fracture.