• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2002.10b
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• pp.195-196
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• 2002

Studied is fretting wear behaviour of transversely vibrating tube which is supported by springs and dimples. This simulates the fuel rod fretting due to flow-induced vibration in a nuclear reactor. The contact between spacer grid springs and fuel cladding tubes arc brought into focus in this paper. From the mechanical viewpoint, a concave contact shape of spring is considered to perform a wider distribution of the contact stress. Sliding/impacting experiments are conducted in air at room temperature with the conditions of positive contact force and gap existence to accommodate the mechanical condition between the fuel rod and the grid spring during reactor operation. It is found that wear region is separated and wear volume becomes larger as the supporting condition becomes poorer. Spring and dimple cause similar wear.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2002.05a
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• pp.304-309
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• 2002

Fretting wear test in room temperature water was performed to evaluate the wear coefficient of Inconel 600,690 (Pressurized Water Reactor, PWR) and Alloy 800 (CANadian DeuteriumUranium, CANDU) steam generator (SG) tubes against ferritic and martensitic stainless steels. The main focus is to compare the wear behaviors between Alloy 800 and Inconel alloys. Test conditions are $10{\sim}30N$ of normal load, $200{\sim}450{\mu}m$ of sliding amplitude and 30Hz of frequency. The result indicated that the wear rate of Alloy 800 was higher than those of Inconel 690 at various test condition such as normal loads, sliding amplitudes etc. From the results of SEM observation, there was little evidence of plastic deformation layer that were dominantly formed on the worn surfaces of Inconel 690. Also, wear particles in Alloy 800 were released from contacting asperities deformed by severe plastic flow during fretting wear. Main cause of wear rate between Alloy 800 and Inconel 690 may be due to the difference of hardness between martensitic and ferritic stainless steel. The wear rate and wear mechanism of two tubes in room temperature water are discussed.

• KSTLE International Journal
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• v.3 no.1
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• pp.68-73
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• 2002

A fretting wear experiment in roam temperature air was performed to evaluate the wear mechanism of fuel rod using a fretting wear tester, which has been developed for experimental study, The main focus was to compare the wear behaviors of fuel rod against support springs with different contact contours (i.e. concave and convex). Wear volume, degree or surface hardening and adhesion tendency of wear particle were examined by the surface roughness tester. The result indicated that with a change of contact condition from contact force of 5 N to 0.1 mm gap, the wear volume of tube increased in the condition of concave spring, but slowly decreased in convex spring. From the results of SEM observation, wear mechanism of each test condition was also dependent on the spring shapes. The wear mechanism of each test condition in room temperature air is discussed.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.2
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• pp.61-68
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• 2023

The phenomenon of fretting wear due to the flow-induced vibration in steam generator (SG) tube is a significant degradation mechanism in nuclear power plants. Fretting wear in SG tube is primarily attributed to the friction and impact forces between the SG tube and the tube support structures, experienced during nuclear power plants operation. While the Archard model has generally been used for the prediction of fretting wear in SG tube, it is limited by its linear nature. In this study, we introduced an "Impact Shear Work-rate" (ISW) model, which takes into account the combined effects of impact and sliding. The ISW model was evaluated using existing experimental data on fretting wear in SG tube and was compared against the Archard model. The prediction results using the ISW model were more accurate than those using the Archard model, particularly for impact forces.

• Tribology and Lubricants
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• v.24 no.6
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• pp.362-367
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• 2008

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. The tube and support materials were Inconel 690 and STS 409. The wear tests were conducted in various environments, which are in water without flow, in flowing water and in flowing water with air. The results showed that the flow of water influenced on the wear-life of tube. The wear-life of tube decreased in water flow as compared with wear-life in stationary water.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2008.11a
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• pp.113-114
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• 2008

• Proceedings of the KSME Conference
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• 2003.11a
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• pp.616-621
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• 2003

Fluid-elastic instability is believed to be a cause of the large-amplitude vibration and resulting rapid wear of heat exchanger tubes when the flow velocity exceeds a critical value. For sub-critical flow velocities, the random turbulence excitation is the main mechanism to be considered in predicting the long-term wear of steam generator tubes. Since flow-induced interactions of the tubes with tube supports in the sub-critical flow velocity can cause a localized tube wear, tube movement in the clearance between the tube and tube support as well as the normal contact force on the tubes by fluid should be maintained as low as possible. A simplified method is used for predicting fretting-wear damage of the double $90^{\circ}$U-bend tubes. The approach employed is based on the straight single-span tube analytical model proposed by Connors, the linear structural dynamic theory of Appendix N-1300 to ASME Section III and the Archard's equation for adhesive wear. Results from the presented method show a similar trend compared with the field data. This method can be utilized to predict the fretting-wear of the double $90^{\circ}$U-bend tubes in steam generators.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.21-26
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• 1996

Recently several PWR Nuclear Plant experienced fuel rod fretting wear failures due to Flow Induced Vibration. When such multi-span supported fuel assembly has vibration excitation, it is important to know how fretting wears are progress and when the fuel rods are start to failure. In this study, we estimate the amount of wear depth using Archard theory when the fuel rod starts to relative motion against spacer grid dimples.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.31 no.10
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• pp.999-1008
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• 2007

The spacer grid set is a component in the nuclear fuel assembly. The set supports the fuel rods safely. Therefore, the spacer grid set should have sufficient strength for the external impact forces such as earthquake. The fretting wear occurs between the spring of the fuel rod and the spacer grid due to flow-induced vibration. Conceptual design of the spacer grid set is performed based on the Independence Axiom of axiomatic design. Two functional requirements are defined for the impact load and the fretting wear, and corresponding design parameters are selected. The overall flow of design is defined according to the application of axiomatic design. Design for the impact load is carried out by using nonlinear dynamic analysis to determine the length of the dimple. Topology optimization is carried out to determine a new configuration of the spring. The fretting wear is reduced by shape optimization using the homology theory. The deformation of a structure is called homologous if a given geometrical relationship holds before, during, and after the deformation. In the design to reduce the fretting wear, the deformed shape of the spring should be the same as that of the fuel rod. This condition is transformed to a function and considered as a constraint in the shape optimization process. The fretting wear is expected to be reduced due to the homology constraint. The objective function is minimizing the maximum stress to allow a slight plastic deformation. Shape optimization results are confirmed through nonlinear static analysis.

• Journal of the Korean Institute of Gas
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• v.4 no.1 s.9
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• pp.26-32
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• 2000

The fretting wear corrosion characteristics between the SM20C and the SM20C, the YBsC3 and the STC4H was experimented by using radical type friction experimental device under the corrosion environment of atmosphere, neutral solution, acid solution and chemical factors of the sea water. The affection of underground water that affect fretting wear corrosion of the SM20C which is moving specimen was more sensitive at the STC4H and more insensible at the YBsC3. The affection of underground water that affect fretting wear corrosion of the STC4H was less, but in the $0.5\%\;H_2SO_4$ and $0.5\%\;HNO_3$ solutions the fretting wear corrosion of the STC4H was more large. The fretting wear corrosion of the SM20C which is moving specimen in the underground water was less than in the $3.5\%\;NaCl$, $0.5\%\;H_2SO_4$ and $0.5\%\;HNO_3$ solutions. As time passed, the fretting wear corrosion is increased in the $HNO_3$ solution and dull in the $0.5\%\;H_2SO_4$ solution.