Fretting Wear Characteristics of Nuclear Fuel Cladding in High-Temperature Pressurized Water

Abstract

Abstract In pressurized water reactor (PWR), fretting wear is one of the main causes of fuel assembly failure. Moreover, the operation condition of cladding is complex and harsh. A unique fretting damage test equipment was developed and tested to simulate the fretting damage evolution process of cladding in the PWR environment. It can simulate the fretting wear experiment of PWR under different temperatures (maximum temperature is 350 ℃), displacement amplitude, vibration frequency, and normal force. The fretting wear behavior of Zr-4 alloy under different temperature environments was tested. In addition, the evolution of wear scar morphology, profile, and wear volume was studied using an optical microscope (OM), scanning electron microscopy (SEM), and a 3D white light interferometer. Results show that higher water temperature evidently decreased the cladding wear volume, the wear mechanism of Zr-4 cladding changed from abrasive wear to adhesive wear and the formation of an oxide layer on the wear scar reduced the wear volume and maximum wear depth.

Keywords

• Fretting wear
• Cladding
• High temperature and high pressure
• Zirconium alloy