Evolution of fretting wear with the number of cycles on Inconel 690 steam generator tubes against AISI 420 steel under gross slip conditions

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.

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Structural Integrity Assessment of Steam Generator Tube by the Use of Heterogeneous Finite Element Method

Journal of Pressure Vessel Technology ◽

10.1115/1.2967727 ◽

2008 ◽

Vol 130 (4) ◽

Cited By ~ 1

Author(s):

Xinjian Duan ◽

Michael J. Kozluk ◽

Sandra Pagan ◽

Brian Mills

Keyword(s):

Finite Element ◽

Steam Generator ◽

Failure Modes ◽

Structural Integrity ◽

Fretting Wear ◽

Defect Depth ◽

Steam Generator Tube ◽

Integrity Assessment ◽

Failure Pressure ◽

Steam Generator Tubes

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.

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Development of Wear Test System for Steam Generator Tubes in Nuclear Power Plants

Key Engineering Materials ◽

10.4028/www.scientific.net/kem.297-300.1418 ◽

2005 ◽

Vol 297-300 ◽

pp. 1418-1423 ◽

Cited By ~ 1

Author(s):

Chi Yong Park ◽

Yong Sung Lee ◽

Myung Hwan Boo

Keyword(s):

Steam Generator ◽

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Wear Test ◽

Test System ◽

Fretting Wear ◽

Stable Operation ◽

Test Results ◽

Steam Generator Tubes

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.

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Fretting Wear Characteristics of INCONEL 690 and INCONEL 600 in Water Environment

Key Engineering Materials ◽

10.4028/www.scientific.net/kem.297-300.1412 ◽

2005 ◽

Vol 297-300 ◽

pp. 1412-1417 ◽

Cited By ~ 8

Author(s):

Sung Hoon Jeong ◽

Chi Yong Park ◽

Young Ze Lee

Keyword(s):

Nuclear Power ◽

Water Environment ◽

Fretting Wear ◽

Wear Loss ◽

Stick Slip ◽

Inconel 600 ◽

Delamination Wear ◽

Slip Regime ◽

Inconel 690 ◽

Steam Generator Tubes

Fretting is the oscillatory motion with very small amplitudes, which usually occurs between two solid surfaces in contact. Fretting wear is the removal of material from contacting surfaces through fretting action. Fretting wear of steam generator tubes in nuclear power plant becomes a serious problem in recent years. The materials for the tubes usually are INCONEL 690 (I-690) and INCONEL 600 (I-600). In this paper, fretting wear tests for I-690 and I-600 were performed under various applied loads in water at room temperature. Results showed that the fretting wear loss of I-690 and I-600 tubes was largely influenced by stick-slip. The fretting wear mechanisms were the abrasive wear in slip regime and the delamination wear in stick regime. Also, I-690 had somewhat better wear resistance than I-600.

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