Wear Characteristics of INCONEL 690 and INCONEL 600 in Elevated Temperature

2005 ◽  
Vol 297-300 ◽  
pp. 1424-1429 ◽  
Author(s):  
Sung Hoon Jeong ◽  
Young Ze Lee
Keyword(s):  
Wear Mechanisms ◽  
Nuclear Power ◽  
Power Plants ◽  
Elevated Temperatures ◽  
Fretting Wear ◽  
Wear Characteristics ◽  
Inconel 600 ◽  
Wear Life ◽  
Inconel 690 ◽  
C 50

In this paper, the fretting wear characteristics of INCONEL 690 (I-690) and INCONEL 600 (I-600) was evaluated to verify the wear mechanism and the wear life. Because of the excellent corrosion-resistance of nickel-based alloy, those materials are used for steam generator tube in nuclear power plants. Sometimes the tubes are damaged due to small amplitude vibration, so called fretting wear. To verify the fretting wear mechanisms the wear experiment was carried with the crossed-cylinder wear tester, which used a cam to oscillate the specimen. The test was carried out at loads of 40N and 90N in elevated temperatures of water. The temperatures of water were 20°C, 50°C and 80°C. The increase of water temperature causes the oxidation of the contact area to be delayed, and the amount of wear on oxide layer to be reduced. The main wear mechanisms of fretting were abrasive wear and oxidation wear.

Fretting Wear Characteristics of Tube-Support Components for a Steam Generator in Elevated Temperature

2005 ◽  
Author(s):  
Sung-Hoon Jeong ◽  
Young-Ze Lee
Keyword(s):  
High Pressure ◽  
Wear Mechanisms ◽  
Nuclear Power ◽  
Fretting Wear ◽  
Wear Volume ◽  
Steam Generators ◽  
Flow Induced Vibration ◽  
Wear Characteristics ◽  
Contact Surfaces ◽  
Contact Part

Tubes in nuclear steam generators are held up by supports because the tubes are long and slender. Fluid of high-pressure and high-temperature flows in the tubes and the flows cause oscillating motions between tubes and supports. This is called FIV (flow induced vibration) which cause fretting wear in contact part of tube-support. The fretting wear of tube-support can threaten the safety of nuclear power plant. Therefore, a research about the fretting wear characteristics of tube-support is required. This work is focused on investigations of fretting wear characteristics and wear mechanisms of tube-support. Results are that the wear rate of tube is proportional to that of support and as water temperature increases the wear volume of tube-support decreases because the oxidation rate decreases due to lack of the oxygen concentration in contact surfaces. Also, the wear mechanisms of tube-support are abrasive and oxidational wear.

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

2005 ◽  
Vol 297-300 ◽  
pp. 1412-1417 ◽  
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.

Wear Characteristics of Tube-Support Components for a Nuclear Steam Generator under Fretting Conditions

2007 ◽  
Vol 120 ◽  
pp. 181-186 ◽  
Author(s):  
Sung Hoon Jeong ◽  
Young Ze Lee
Keyword(s):  
Wear Mechanisms ◽  
Nuclear Power ◽  
Fluid Flows ◽  
Fretting Wear ◽  
Wear Volume ◽  
Steam Generators ◽  
Flow Induced Vibration ◽  
Wear Characteristics ◽  
Contact Surfaces ◽  
Contact Part

Tubes in nuclear steam generators are held up by supports because the tubes are long and slender. Fluid flows of high-pressure and high-temperature flows in the tubes cause oscillating motions between tubes and supports. This is called as FIV (flow induced vibration) which cause fretting wear in contact part of tube-support. The reduction of tube thickness due to fretting wear of tube-support can threaten the safety of nuclear power plant. Therefore, a research on the fretting wear characteristics of tube-support is required. This work is focused on investigations of fretting wear characteristics and wear mechanisms of tube-support. Results show that the wear rate of tube is proportional to that of support and that with increasing the water temperature the wear volume of tube-support decreases because the oxidation rate decreases due to reduction of the oxygen concentration in contact surfaces. Also, the wear mechanisms of tube-support are abrasive and oxidational wear.

Study on the Fretting Wear Behavior and Mechanism of Nuclear Alloy 690 Tube

2012 ◽  
Vol 512-515 ◽  
pp. 1740-1746
Author(s):  
Jin Na Mei ◽  
Fei Xue ◽  
Zhao Xi Wang ◽  
Guo Dong Zhang ◽  
Lei Huang ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Wear Behavior ◽  
Damage Mechanism ◽  
Fretting Wear ◽  
Composition Analysis ◽  
Ambient Conditions ◽  
Inconel 600 ◽  
Delamination Wear ◽  
Slip Regime ◽  
Inconel 690

The fretting wear behavior of Inconel 690 for steam generator tube in a nuclear power plant against Inconel 600 (Cr plating) for anti-vibration bar was investigated in a plain cylinder/plane contact configuration under ambient conditions. The fretting wear tests were conducted under various applied normal loads of 20-100 N, slip amplitudes of 20-100 μm and frequency of 2 Hz. Observation of worn surface and corresponding chemical composition analysis were performed to clarify the fretting wear mechanism. It is found that the friction coefficient value increases firstly and then tends to stabilize with decrease in normal loads and increase in slip amplitudes. In addition, the fretting regime is identified to be gross slip regime, indicating that the fretting damage mechanism of Inconel 690 tube against Inconel 600 (Cr plating) bar is wear. The corresponding fretting wear mechanisms are dominant by delamination wear, abrasive wear and friction oxidation.

The Effects of Ion Implantation on the Surface Features of Inconel 600

1985 ◽  
Vol 43 ◽  
pp. 288-289
Author(s):  
John D. Rubio
Keyword(s):  
Grain Boundary ◽  
Ion Implantation ◽  
Nuclear Power ◽  
Power Plants ◽  
Surface Region ◽  
Selective Dissolution ◽  
Boundary Region ◽  
Near Surface ◽  
Inconel 600 ◽  
Steam Generator Tubing

The degradation of steam generator tubing at nuclear power plants has become an important problem for the electric utilities generating nuclear power. The material used for the tubing, Inconel 600, has been found to be succeptible to intergranular attack (IGA). IGA is the selective dissolution of material along its grain boundaries. The author believes that the sensitivity of Inconel 600 to IGA can be minimized by homogenizing the near-surface region using ion implantation. The collisions between the implanted ions and the atoms in the grain boundary region would displace the atoms and thus effectively smear the grain boundary.To determine the validity of this hypothesis, an Inconel 600 sample was implanted with 100kV N2+ ions to a dose of 1x1016 ions/cm2 and electrolytically etched in a 5% Nital solution at 5V for 20 seconds. The etched sample was then examined using a JEOL JSM25S scanning electron microscope.

Wear Transitions of Tube-Support Components for a Nuclear Steam Generator under Fretting Conditions

2006 ◽  
Vol 326-328 ◽  
pp. 1263-1266 ◽  
Author(s):  
Sung Hoon Jeong ◽  
Jung Min Park ◽  
Joong Hui Lee ◽  
Young Ze Lee
Keyword(s):  
Nuclear Power ◽  
Applied Load ◽  
Fluid Flows ◽  
Relative Displacement ◽  
Fretting Wear ◽  
Steam Generators ◽  
Flow Induced Vibration ◽  
Support Materials ◽  
Mild Wear ◽  
Inconel 690

Tubes in nuclear steam generators are held up by supports because the tubes are long and slender. Fluid flows of high-pressure and high-temperature in the tubes cause oscillating motions between tubes and supports. This is called as FIV (flow induced vibration), which causes fretting wear in contact parts of tube-support. The fretting wear of tube-support can threaten the safety of nuclear power plant. Therefore, a research about the fretting wear characteristics of tube-support is required. This work is focused on fretting wear transitions from mild wear to severe wear of tube-support materials by various loads and relative displacements. The transition is defined on the basis of the changes in wear amount. To investigate the transition, the fretting wear tester was contrived to prevent the reduction of relative displacement between tube and support by increasing the load. The tube and support materials were Inconel 690 and 409 SS, respectively. The results show that the transition of tube-support wear is caused by the changes of the dominant wear mechanism depending on the applied load and the relative displacement.

Fretting wear mechanisms of Zircaloy-4 and Inconel 600 contact in air

2001 ◽  
Vol 15 (9) ◽  
pp. 1274-1280 ◽  
Author(s):  
Tae-Hyung Kim ◽  
Seock-Sam Kim
Keyword(s):  
Wear Mechanisms ◽  
Fretting Wear ◽  
Inconel 600

The Effect of Elevated Temperatures and Nuclear Radiation on the Properties of Biological Shielding Concrete

2016 ◽  
Vol 677 ◽  
pp. 8-16 ◽  
Author(s):  
Jaroslava Koťátková ◽  
Jan Zatloukal ◽  
Pavel Reiterman ◽  
Jan Patera ◽  
Zbyněk Hlaváč ◽  
...  
Keyword(s):  
Gamma Radiation ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Elevated Temperatures ◽  
Nuclear Radiation ◽  
Containment Vessel ◽  
Negative Impacts

The paper reviews the so far known information about the properties of biological shielding concrete used in the containment vessel of nuclear power plants (NPP) and its behaviour when exposed to radiation. The damage of concrete caused by neutron and gamma radiation as well as by the accompanying generation of heat is described. However, there is not enough data for the proper evaluation of the negative impacts and further research is needed.

scholarly journals 14C Measurements at PWR-Type Nuclear Power Plants in Three Middle European Countries

Radiocarbon ◽  
1997 ◽  
Vol 40 (1) ◽  
pp. 439-446 ◽  
Author(s):  
György Uchrin ◽  
Ede Hertelendi ◽  
Gábor Volent ◽  
Ondrej Slavik ◽  
Jozef Morávek ◽  
...  
Keyword(s):  
Power Unit ◽  
Nuclear Power ◽  
Power Plants ◽  
Discharge Rate ◽  
Continuous Increase ◽  
Mean Values ◽  
A Value ◽  
Discharge Rates ◽  
C 50 ◽  
Ring Analysis

Regular 14C sampling of discharged air began in 1988 at Paks Nuclear Power Plant (NPP), Hungary, and in 1991 at NPPs in Krsko, Slovenia and Bohunice, Slovakia. Monitoring of 14C discharges is carried out at all NPPs with similar differential samplers continuously collecting 14C in the form of 14CO2 and 14CnHm. The main results of airborne discharge monitoring are as follows: 14C activity concentration varied roughly within a factor of two around their mean values, 125 Bq m-3 and 90 Bq m-3 for Paks and for Krsko NPP, respectively. The pattern of discharge for Bohunice NPP is slightly different from that at the other two stations. At Bohunice, there has been a continuous increase in the discharge rate at power unit V1, starting with 70 Bq m-3 in 1991 and reaching a value of 190 Bq m-3 in 1995. The values for power unit V2 are 50 Bq m-3 and 82 Bq m-3. The average normalized yearly discharge rates are 0.887 (TBqGWe-1yr-1) for Paks, 0.815 (V1) and 0.500 (V2) for Bohunice, and 0.219 for Krsko. Most of the discharged 14C is in hydrocarbon form, 95% for Paks and Bohunice V2, but the CO2 fraction may reach 25% or 43% at Bohunice V1 and Krsko, respectively. At Bohunice V1, not only the discharge rate increased but the 14CO2 ratio to the total changed from 30% to 13%. The local radiological impact is estimated to be 1.5 μSv a-1 for Paks, 1.7 μSv a-1 for Bohunice, and 0.12 μSv a-1 for Krsko. The 14C excess in the environment has been measured at Paks NPP since 1989. Based on the monitoring data, the long-term average 14C excess from the Paks NPP was D14C=50% for hydrocarbons. Tree-ring analysis has shown a slight excess around Krsko NPP: D14C is equal to 199.9% for a tree at 1 km from the NPP compared with a “reference” one for which D14C was equal to 111.6% (in 1994).

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