Micro-amplitude impact wear behavior of 690 alloy under deionized water and dry conditions

被引:0
作者
Cai, Zhen-Bing [1 ]
Deng, Xiao-Jian [1 ]
Yang, Rong [1 ]
Peng, Jin-Fang [1 ]
Qian, Hao [2 ]
Li, Chen [2 ]
Xie, Yong-Cheng [2 ]
Zhu, Min-Hao [1 ]
机构
[1] Tribology Research Institute, Key Laboratory of Advanced Materials Technology, Southwest Jiaotong University, Chengdu
[2] Shanghai Nuclear Engineering Research and Design Institute, Shanghai
来源
Zhendong yu Chongji/Journal of Vibration and Shock | 2015年 / 34卷 / 04期
关键词
690; alloys; Impact wear; Lubrication conditions; Wear mechanism;
D O I
10.13465/j.cnki.jvs.2015.04.003
中图分类号
学科分类号
摘要
Through the innovation of fixtures in a small load testing machine for impact wear, the contact between a tube and a plane was realized to simulate the micro-amplitude impact wear between a tube and an anti-vibration bar in a steam generator. Here, a 690 alloy tube and 405 stainless steel were selected as the test materials. Micro-amplitude impact wear tests were conducted with different impact loads and lubrication conditions (dry and deionized water).The wear and damage mechanisms of 690 alloy were investigated through analyzing polishing scratch area, micro-topography, fracture morphology, and micro hardness, etc. The results showed that the wear mechanism of 690 alloy under the small load and dry conditions is oxidation it is oxidation and fatigue wear under growing loads; under deionized water, a slight damage can be found with small loads, and the wear mechanism is fatigue wear under growing loads; water liquid delays the cracking time significantly. ©, 2015, Chinese Vibration Engineering Society. All right reserved.
引用
收藏
页码:14 / 18
页数:4
相关论文
共 12 条
[1]  
Fujita K., Flow-induced vibration and fluid-structure interaction in nuclear power plant components, Journal of Wind Engineering and Industrial Aerodynamics, 33, pp. 405-418, (1990)
[2]  
Ko P.L., Wear of power plant components due to impact and sliding, Appl Mech Rev, 50, 7, pp. 387-411, (1997)
[3]  
Goyder H.G.D., Flow-induced vibration in heat exchangers, Chemical Engineering Research and Design, 80, 3, pp. 226-232, (2002)
[4]  
Ko P.L., Lina A., Ambard A., A review of wear scar patterns of nuclear power plant components, ASME 2003 Pressure Vesselsand Piping Conference, pp. 97-106, (2003)
[5]  
Reynier B., Phalippou C., Riberty P., Et al., Influence of a periodic latency time on the impact/sliding wear damage of two PWR control rods and guide cards specimens, Wear, 259, 7-12, pp. 1314-1323, (2005)
[6]  
Pagan S., Duan X., Kozluk M.J., Et al., Characterization and structural integrity tests of ex-service steam generator tubes at Ontario Power Generation, Nuclear Engineering and Design, 239, 3, pp. 477-483, (2009)
[7]  
Ding C.-Y., Shen S.-F., Jia D.-N., The analysis and experimental study of impacting force of the tube against support plate in heat exchangers, Nuclear and Science, 9, 1, pp. 34-44, (1989)
[8]  
Wang T.-H., Shen S.-F., Experimental studies of fretting wear in heat exchanger tubes, Nuclear Power Engineering, 11, 4, pp. 338-443, (1990)
[9]  
Ding X.-S., Fretting wear and protection of steam generator tubes, Nuclear Safe, 3, pp. 27-32, (2006)
[10]  
Wang Y., Shi X.-Y., Cai L.-J., Et al., Development of a new small load testing machine for impact wear and testing of its characteristics, Tribology, 27, 5, pp. 487-491, (2007)