Sliding Wear Behaviors of Steam Generator Tubes in Various Environments

2006 ◽  
Vol 510-511 ◽  
pp. 566-569
Author(s):  
Gyung Guk Kim ◽  
Seung Dae Noh ◽  
Gi Sung Park ◽  
Seon Jin Kim ◽  
Deok Hyun Lee ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Sliding Wear ◽  
Power Plants ◽  
Wear Behavior ◽  
Water Environment ◽  
Wear Volume ◽  
Wear Properties ◽  
Inconel 600 ◽  
Steam Generator Tubes

Wear damage of steam generator tubes for nuclear power plants can cause the leakage of radioactive substances. Therefore, the evaluation of integrity and safety for tubes is very important from the viewpoint of nuclear ecocide. In the present study, to investigate the wear properties of Inconel 600 and 690 steam generator tube materials mated with 409 stainless steel commonly used as support plate, sliding wear tests were performed with increasing sliding distance in air and in elevated temperature water environment, respectively. The wear volume of tube materials was less than those of supports under all conditions. There were no significant differences in the wear behavior for the Inconel 600 and 690 tubes, independently of the testing environment.

Sliding Wear Behavior of Steam Generator Tube Materials in Nuclear Power Plants

2005 ◽  
Vol 486-487 ◽  
pp. 137-140 ◽  
Author(s):  
Gyung Guk Kim ◽  
Ji Hui Kim ◽  
Kwon Yeong Lee ◽  
Seon Jin Kim ◽  
Deok Hyun Lee ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Steam Generator ◽  
Nuclear Power ◽  
Sliding Wear ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Wear Behavior ◽  
Steam Generator Tube ◽  
Inconel 600 ◽  
Steam Generator Tubes

Wear damage of steam generator tubes of nuclear power plants can cause leakage of radioactive substances. So the evaluation of tubes’ integrity is very important from the viewpoint of nuclear ecocide. In the present study, sliding wear behaviors of Inconel 600 and 690 steam generator tube materials mated with 409 stainless steel commonly used as the support plate were investigated at room temperature in an air environment. For more precise prediction of wear behaviors of steam generator tubes, Archard equation was modified, and the modified wear coefficients were estimated as a function of sliding distance. When using the modified Archard equation, the reliabilities for prediction of wear behavior of Inconel 600 and 690 mated with 409 stainless steel increased from 71.8% to 83.8% and from 60.2% to 85.2%, respectively.

scholarly journals Wear Induced by Stochastic Sliding Impacts

2015 ◽  
Author(s):  
Thibaut Souilliart ◽  
Emmanuel Rigaud ◽  
Alain Le Bot ◽  
Christian Phalippou
Keyword(s):  
Impact Velocity ◽  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Incidence Angle ◽  
Wear Volume ◽  
Test Rig ◽  
Normal And Tangential Loads ◽  
Steam Generator Tubes ◽  
Impact Duration

Vibrations of the steam generator tubes in nuclear power plants induce stochastic impacts between the tubes and their supports. As a consequence, wear is generated. A test rig is designed and used to perform impacts between two metal crossed cylinders with various incidence angles and impact velocities. The normal and tangential components of the contact load are measured during the tests. Rate and duration of impacts, instantaneous ratio between normal and tangential loads for each impact are deduced. Influence of incidence angle and impact velocity on impact duration, ratio between tangential and normal loads during impact and wear volume is highlighted.

Reciprocating Sliding Wear Behavior of Inconel 600 Mated with SUS 304 Stainless Steel at Elevated Temperatures

2006 ◽  
Vol 510-511 ◽  
pp. 270-273
Author(s):  
Kwon Yeong Lee ◽  
Chang Seon Hwang ◽  
Seon Jin Kim ◽  
Gyung Guk Kim ◽  
Ji Hui Kim ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Sliding Wear ◽  
Elevated Temperatures ◽  
Wear Behavior ◽  
304 Stainless Steel ◽  
Wear Volume ◽  
Wear Coefficient ◽  
Inconel 600 ◽  
Error Range ◽  
Sliding Distance

The sliding wear behavior of a steam generator in a nuclear power plant (Inconel 600) was investigated at 225, 250 and 300°C. Effects of wear parameters such as sliding distance and contact stress were examined with SUS 304 (austenitic). In the prediction of the wear volume by Archard wear equation, the standard error range was calculated to be ±0.53×10-9 m3 and the reliability to be 71.9 % for the combination of Inconel 600 and SUS 304. The error range was considered to be relatively broad because the wear coefficient in Archard equation was assumed to be constant. However, the wear volume turned out to increase parabolically with the sliding distance owing to the combination of strain hardening of the wear surface and the oxidative wear. Based on the experimental results, the wear coefficient was modified as a function of the rotating sliding distance. The calculation with the modified wear equation showed that, although the error range was not significantly narrowed, the reliability increased from 71.9 to 78.1 %.

Sliding Wear Behavior of Steam Generator Tube Materials in Nuclear Power Plants

2005 ◽  
pp. 137-140
Author(s):  
Gyung Guk Kim ◽  
Ji Hui Kim ◽  
Kwon Yeong Lee ◽  
Seon Jin Kim ◽  
Deok Hyun Lee ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Sliding Wear ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Wear Behavior ◽  
Steam Generator Tube

The Geometry of Steam Generator Tube and Its Relevance to the Occurrence of Stress Corrosion Cracking in Operating Nuclear Power Plants

2010 ◽  
Author(s):  
Deok Hyun Lee ◽  
Do Haeng Hur ◽  
Myung Sik Choi ◽  
Kyung Mo Kim ◽  
Jung Ho Han ◽  
...  
Keyword(s):  
Stress Corrosion ◽  
Steam Generator ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Corrosion Cracking ◽  
Location Type ◽  
Steam Generator Tubes ◽  
Non Destructive

Occurrences of a stress corrosion cracking in the steam generator tubes of operating nuclear power plants are closely related to the residual stress existing in the local region of a geometric change, that is, expansion transition, u-bend, ding, dent, bulge, etc. Therefore, information on the location, type and quantitative size of a geometric anomaly existing in a tube is a prerequisite to the activity of a non destructive inspection for an alert detection of an earlier crack and the prediction of a further crack evolution [1].

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