Fretting Wear Damage of the Corroded Fuel Cladding Tubes for Nuclear Fuel Rod against Supporting Grids

2007 ◽  
Vol 345-346 ◽  
pp. 709-712
Author(s):  
Jin Seon Kim ◽  
Yong Hwan Kim ◽  
Seung Jae Lee ◽  
Young Ze Lee
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Fluid Flows ◽  
Fretting Wear ◽  
Wear Depth ◽  
Fuel Cladding ◽  
Flow Induced Vibration ◽  
Nuclear Fuel Assembly ◽  
Wear Damage ◽  
Increasing Load

Fuel cladding tubes in nuclear fuel assembly are held up by supporting grids 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. The fretting wear tests were performed with supporting grids and cladding tubes, especially after corrosion treatment on tubes, in water. The tests were done using various applied loads with fixed amplitude. From the results of fretting tests, the wear amounts of tube materials can be predictable by obtaining the wear coefficient using the work rate model. Due to stick phenomena the wear depth was changed as increasing load and temperature. The maximum wear depth was decreased as increasing the water temperatures. At high temperatures there are the regions of some severe adhesion due to stick phenomena.

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.

A Numerical Characterization of Flow-Induced Vibration and Fretting Wear Potential in Nuclear Steam Generator Tube Bundles

2011 ◽  
Author(s):  
Marwan Hassan ◽  
Atef Mohany
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Complex Dynamics ◽  
Fretting Wear ◽  
Contact Ratio ◽  
Flow Induced Vibration ◽  
Numerical Characterization ◽  
Wear Damage ◽  
Effect Of Support ◽  
Tube Failure

Nuclear power plants have experienced problems related to tube failures in steam generators. While many of these failures have been attributed to corrosion, it has been recognized that flow-induced vibrations contribute significantly to tube failure. In order to avoid these excessive vibrations, tubes are stiffened by placing supports along their length. Various tube/support geometries have been used, but the majority are either support plates (plates with drilled or broached holes) or flat bars. Unfortunately, clearance is often considered necessary between the tubes and their supports to facilitate tube/support assembly and to allow for thermal expansion of the tubes. A combination of flow-induced turbulence and fluidelastic forces may then lead to unacceptable tube fretting-wear at the supports. The fretting wear damage could ultimately cause tube failure. Such failures may require shut downs resulting in production losses, and pose potential threats to human safety and the environment. Therefore, it is imperative to predict the nonlinear tube response and the associated fretting wear damage to tubes due to fluid excitations. Tubes in loose flat-bar supports have complex dynamics due to the possible combinations of geometry. The understanding of tube dynamics in the presence of this type of support and the associated fretting wear is still incomplete. These issues are addressed in this paper through simulations of the dynamics of tubes subjected to crossflow turbulence and fluidelastic instability forces. The finite element method is utilized to model the vibrations and impact dynamics. The tube model simulates a U-tube supported by 16 flat bars with clearances and axial offset. Results are presented and comparisons are made for the parameters influencing the fretting-wear damage such as contact ratio, impact forces and normal work rate. The effect of support clearance and support axial offset are investigated.

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.

Wear Transitions of Tube-Support Materials for a Nuclear Steam Generator through Sliding Wear Test and Fretting Wear Test

2006 ◽  
Vol 321-323 ◽  
pp. 430-433 ◽  
Author(s):  
Sung Hoon Jeong ◽  
Byoung Jong Lee ◽  
Young Ze Lee
Keyword(s):  
Nuclear Power ◽  
Fluid Flows ◽  
Wear Test ◽  
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.

Influence of Temperature on the Fretting Wear of Advanced Nuclear Fuel Cladding Tube against Supporting Grid

2007 ◽  
Vol 345-346 ◽  
pp. 705-708 ◽  
Author(s):  
Young Chang Park ◽  
Sung Hoon Jeong ◽  
Yong Hwan Kim ◽  
Seung Jae Lee ◽  
Young Ze Lee
Keyword(s):  
Water Temperature ◽  
Oxide Layer ◽  
Nuclear Fuel ◽  
Wear Mechanism ◽  
Nuclear Reactor ◽  
Fretting Wear ◽  
Wear Volume ◽  
Wear Depth ◽  
Flow Induced Vibration ◽  
Worn Surface

The experimental investigation was performed to find the associated changes in characteristics of fretting wear with various water temperatures. Fretting can be defined as the oscillatory motion with very small amplitudes, which usually occur between two contacting surfaces. The fretting wear, which occurs between cladding tubes of nuclear fuel rod and grids, causes in damages the cladding tubes by flow induced vibration in a nuclear reactor. In this paper, the fretting wear tests were carried out using the zirconium alloy tubes and the grids with increasing the water temperature. The tube materials in water of 20, 50 and 80 were tested with the applied load of 20N and the relative amplitude of 200. The worn surfaces were observed by SEM, EDX and 2D surface profiler. As the water temperature increased, the wear volume was decreased. However, oxide layer was increased on the worn surface. The abrasive wear mechanism was observed at water temperature of 20 and adhesive wear mechanism occurred at water temperature of 50 and 80. As the water temperature increased, surface micro-hardness was decreased. Also, wear depth and wear width were decreased due to increasing stick phenomenon. Stick regime occurred due to the formation of oxide layer on the worn surface with increasing water temperatures.

An Evaluation of Fuel Rod Fretting Wear in Spacer Grid With Conformal Shape

2006 ◽  
Author(s):  
Young Ki Jang ◽  
Nam Kyu Park ◽  
Jae Ik Kim ◽  
Kyu Tae Kim ◽  
Chong Chul Lee ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Calculation Method ◽  
Power Plants ◽  
Failure Time ◽  
Fretting Wear ◽  
Measuring System ◽  
Wear Volume ◽  
Wear Depth ◽  
Volume Calculation ◽  
Fuel Rod

Turbulent flow-induced vibration in nuclear fuel may cause fretting wear of fuel rod at grid support locations. An advanced nuclear fuel for Korean PWR standard nuclear power plants (KSNPs), has been developed to get higher performance comparing to the current fuel considering the safety and economy. One of the significant features of the advanced fuel is the conformal shape in mid grid springs and dimples, which are developed to diminish the fretting wear failure. Long-term hydraulic tests have been performed to evaluate the fretting wear of the fuel rod with the conformal springs and dimples. Wear volume is a measure to predict the fretting wear performance. The shapes of a lot of scars are non-uniform such as wedge shapes, and axially non-symmetric shapes, etc., depending on the contact angle between fuel rod and springs/dimples. In addition, conformal springs and dimples make wear scars wide and thin comparing to conventional ones with convex shape. It is found that wear volumes of these kinds of non-uniform wear scars are over-predicted when the traditionally used wear depth-dependent volume calculation method is employed. In order to predict wear volume more accurately, therefore, the measuring system with high accuracy has been used and verified by the known wear volumes of standard specimens. The wear volumes of the various wear scars have been generated by the measuring system and used for predicting the fretting wear-induced failure time. Based on new evaluation method, it is considered that the fretting wear-induced fuel failure duration with this conformal grid has increased up to 8 times compared to the traditionally used wear depth-dependent volume calculation method.

Robust design of fuel rods against potential structural damage by flow-induced vibration

2019 ◽  
Vol 33 (01n03) ◽  
pp. 1940008
Author(s):  
Huan-Huan Qi ◽  
Nai-Bin Jiang ◽  
Yi-Xiong Zhang ◽  
Zhi-Peng Feng ◽  
Xuan Huang
Keyword(s):  
Structural Damage ◽  
Flow Distribution ◽  
Fretting Wear ◽  
Leaf Spring ◽  
Wear Depth ◽  
Pressurized Water Reactor ◽  
Flow Induced Vibration ◽  
Vibration Displacement ◽  
Pressurized Water ◽  
Fuel Rod

We studied the flow-induced vibration (FIV) and fretting wear of fuel rod with grid relaxation. According to the flow distribution around a type of pressurized water reactor (PWR) fuel rod, the power spectral density (PSD) is obtained to characterize the turbulence excitation. By combining the correlation of PSD test parameters, the mean square value of the vibration displacement of each rod mode is found, and then the wear depth of dimple position is calculated based on the ARCHARD wear formula. The grids may relax due to inaccurate manufacturing, fuel transportation and in-core irradiation. The absence of grid clamping force would significantly influence the rod mode and thereby changes its FIV responses. Simulation results show that the failure of the leaf spring has negligible effect on the rod natural frequency whereas the dimple failure near the location with larger FIV amplitude has a much significant effect. The lateral flow velocities at the inlet and outlet of the core are larger. For the fully clamped fuel rod, the responses amplitude of turbulent excitation at the bottom and top of the fuel rod are larger. This is even more obvious with a failed dimple at these locations. Comparatively, the effect of dimple support failure in the middle is less influential. The influence of dimple support failure on the rod wear depth depicts basically the same trend as on the maximum FIV amplitude.

Experimental Fretting-Wear Studies of Steam Generator Materials

1995 ◽  
Vol 117 (4) ◽  
pp. 312-320 ◽  
Author(s):  
N. J. Fisher ◽  
A. B. Chow ◽  
M. K. Weckwerth
Keyword(s):  
Steam Generator ◽  
Contact Force ◽  
Environmental Conditions ◽  
Analytical Techniques ◽  
Work Rate ◽  
Fretting Wear ◽  
Flow Induced Vibration ◽  
Dynamic Interactions ◽  
Support Materials ◽  
Wear Damage

Flow-induced vibration of steam generator tubes results in fretting-wear damage due to impacting and rubbing of the tubes against their supports. This damage can be predicted by computing tube response to flow-induced excitation forces using analytical techniques, and then relating this response to resultant wear damage using experimentally derived wear coefficients. Fretting-wear of steam generator materials has been studied experimentally at Chalk River Laboratories for two decades. Tests are conducted in machines that simulate steam generator environmental conditions and tube-to-support dynamic interactions. Different tube and support materials, tube-to-support clearances, and tube support geometries have been studied. The effect of environmental conditions, such as temperature, oxygen content, pH and chemistry control additive, have been investigated as well. Early studies showed that damage was related to contact force as long as other parameters, such as geometry and motion, were held constant. Later studies have shown that damage is related to a parameter called work-rate, which combines both contact force and sliding distance. Results of short and long-term fretting-wear tests for CANDU steam generator materials at realistic environmental conditions are presented. These results demonstrate that work-rate is an appropriate correlating parameter for impact-sliding interaction.

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