Experimental investigation of long-term fretting wear of multi-span steam generator tubes with U-bend sections

Wear ◽  
1999 ◽  
Vol 225-229 ◽  
pp. 563-574 ◽  
Author(s):  
M.H Attia ◽  
E Magel
Wear ◽  
2003 ◽  
Vol 255 (7-12) ◽  
pp. 1198-1208 ◽  
Author(s):  
Young-Ho Lee ◽  
Hyung-Kyu Kim ◽  
Hong-Deok Kim ◽  
Chi-Yong Park ◽  
In-Sup Kim

2020 ◽  
Vol 144 ◽  
pp. 107556 ◽  
Author(s):  
Xianglong Guo ◽  
Ping Lai ◽  
Ling Li ◽  
Lichen Tang ◽  
Lefu Zhang

2006 ◽  
Vol 326-328 ◽  
pp. 1251-1254 ◽  
Author(s):  
Chi Yong Park ◽  
Jeong Keun Lee

Fretting wear generated by flow induced vibration is one of the important degradation mechanisms of steam generator tubes in the nuclear power plants. Understanding of tube wear characteristics is very important to keep the integrity of the steam generator tubes to secure the safety of the nuclear power plants. Experimental examination has been performed for the purpose of investigating the impact fretting. Test material is alloy 690 tube and 409 stainless steel tube supports. From the results of experiments, wear scar progression is investigated in the case of impact-fretting wear test of steam generator tubes under plant operating conditions such as pressure of 15MPa, high temperature of 290C and low dissolved oxygen. Hammer imprint that is actual damaged wear pattern, has been observed on the worn surface. From investigation of wear scar pattern, wear mechanism was initially the delamination wear due to cracking the hard oxide film and finally transferred to the stable impact-fretting pattern.


2008 ◽  
Vol 130 (4) ◽  
Author(s):  
Xinjian Duan ◽  
Michael J. Kozluk ◽  
Sandra Pagan ◽  
Brian Mills

Aging steam generator tubes have been experiencing a variety of degradations such as pitting, fretting wear, erosion-corrosion, thinning, cracking, and denting. To assist with steam generator life cycle management, some defect-specific flaw models have been developed from burst pressure testing results. In this work, an alternative approach; heterogeneous finite element model (HFEM), is explored. The HFEM is first validated by comparing the predicted failure modes and failure pressure with experimental measurements of several tubes. Several issues related to the finite element analyses such as temporal convergence, mesh size effect, and the determination of critical failure parameters are detailed. The HFEM is then applied to predict the failure pressure for use in a fitness-for-service condition monitoring assessment of one removed steam generator tube. HFEM not only calculates the correct failure pressure for a variety of defects, but also predicts the correct change of failure mode. The Taguchi experimental design method is also applied to prioritize the flaw dimensions that affect the integrity of degraded steam generator tubes such as the defect length, depth, and width. It has been shown that the defect depth is the dominant parameter controlling the failure pressure. The failure pressure varies almost linearly with defect depth when the defect length is greater than two times the tube diameter. An axial slot specific flaw model is finally developed.


2005 ◽  
Vol 297-300 ◽  
pp. 1418-1423 ◽  
Author(s):  
Chi Yong Park ◽  
Yong Sung Lee ◽  
Myung Hwan Boo

In steam generators of nuclear power plants, flow-induced vibration (FIV) can lead to tube damage by fretting-wear occurred due to impact and sliding movement between the tubes and their supports. There have been many studies and test results on wear damage of steam generator tubes but they were not reflected the mechanical and chemical conditions accurately. KEPRI nuclear power laboratory developed a wear test system, which is able to control the motion of impact and sliding simultaneously in the pressurized high temperature water-chemistry conditions. Some wear tests were performed to verify the stable operation for the wear test. This wear test system with new concepts was described briefly, and some data for verifying its performance have been shown in the cases of the selected some test results. In the test, Alloy 690 was used for tube materials and 409 stainless steel for support plates. A little data deviation was obtained and stability of system operation was investigated.


Author(s):  
V. P. Janzen ◽  
Y. Han ◽  
B. A. W. Smith ◽  
S. M. Fluit

The integrity of steam-generator tubes is an important aspect of the long-term reliable operation of nuclear power plants. In situations where a tube is judged to be at risk, it must be either plugged, or removed, or reliably stabilized in some manner to avoid excessive motion of the tube due to flow-induced vibration. The present work describes measurements of the effect of an internal cable-type stabilizer on the structural damping of steam-generator tubes. The free-vibration response of unstabilized and stabilized tubes was analyzed to provide damping ratios from frequency-domain spectral responses, time-domain logarithmic decrement ratios and time-domain vibration decay-curves. The structural damping ratios typically increased from approximately 1.6% to approximately 4.3% with the addition of the stabilizer. This last value is somewhat less than recently published values for stabilized tubes from a different type of steam generator, suggesting that tube stabilization, while effective, has limitations that need to be conservatively assessed.


2010 ◽  
Vol 47 (5) ◽  
pp. 449-456 ◽  
Author(s):  
Ki-Wahn RYU ◽  
Chi-Yong PARK ◽  
Hyung-Nam KIM ◽  
Huinam RHEE

Sign in / Sign up

Export Citation Format

Share Document