Preliminary Study on the Fretting Wear Reliability of an Annular Nuclear Fuel

2012 ◽  
Vol 706-709 ◽  
pp. 2535-2539
Author(s):  
Young Ho Lee ◽  
Hyung Kyu Kim
Keyword(s):  
Nuclear Fuel ◽  
Wear Behavior ◽  
Fretting Wear ◽  
Test Results ◽  
Flow Induced Vibration ◽  
Primary Coolant ◽  
Spacer Grids ◽  
Safety Margins ◽  
Power Densities ◽  
Current Operating

Recently, a dual-cooled fuel (i.e. annular fuel) which is compatible with current operating PWR plants has been proposed in order to increase both power densities and safety margins. Due to the design concept that is compatible with current PWR plants, however, when compared with a current solid nuclear fuel it shows a narrow gap between fuel rods and needs to modify spacer grid shapes and their positions. Because a flow-induced vibration by fast primary coolant is inevitable phenomenon, it is necessary to examine the fretting wear behavior between an annular fuel and designed spacer grids. In this study, fretting wear has been performed to evaluate the wear resistance of the annular fuel by using specially designed spring and dimple of spacer grids that have a cantilever type and a hemispherical shape, respectively. At the spring specimen with relatively small stiffness value, fretting wear was initiated at both end regions and then proceeded gradually to center region. Based on the test results, the fretting wear behavior of annular fuel was compared with the current solid nuclear fuel and a comparative factor of its reliability was proposed.

Flow Induced Vibration and Fretting Wear: An Integrated Approach

2002 ◽  
Author(s):  
D. V. Paramonov ◽  
S. J. King ◽  
M. Y. Young ◽  
R. Y. Lu
Keyword(s):  
Flow Field ◽  
Fuel Assembly ◽  
Nuclear Fuel ◽  
Wear Behavior ◽  
Vibration Response ◽  
Fretting Wear ◽  
Analysis Model ◽  
Fuel Rod ◽  
Safety Margins ◽  
Test Loop

Fuel assemblies are exposed to severe thermal, mechanical and radiation loads during operation. Global core and local fuel assembly flow fields typically result in fuel rod vibration. Under certain conditions, this vibration, when coupled with other factors, might result in excessive cladding fretting wear. This phenomenon is of the concern for nuclear fuel designers, especially in light of the need for higher burnup, longer cycle lengths, and operational safety margins in fuel designs. Understanding of (1) the fretting wear margins for a particular nuclear fuel design, (2) the probability of a fuel assembly exposed to a particular set of thermal, mechanical, flow and radiation conditions being at risk of excessive wear, and (3) the factors affecting fretting wear resistance, are important in order to better guide design, testing, and operational flexibility. In this paper, an integrated method to estimate fretting margin of nuclear fuel is presented, including its formulation, benchmark against experimental data and example application to in-core conditions. The major features of the method are as follows: • flow and rod vibration response are coupled through a linear structural analysis model, • flow field is determined using a sub-channel thermal-hydraulic code, • wear progression is treated as a time-dependent process, through taking into account impact of resulting rod-to-support clearance, • a possibility of a fluid-elastic instability is accounted for. Supporting data on basic wear mechanisms, flow field and fuel assembly fretting wear behavior obtained at a number of experimental facilities at Westinghouse Electric Company and Atomic Energy of Canada Limited are also presented. These facility include: • VIPER hydraulic test loop data where vibration response and wear are measured under prototypical flow conditions, and • autoclave fretting-wear machine steam employed to determine fretting-wear coefficients of fuel rod and grid-support designs.

Fretting wear behavior of a nuclear fuel rod under a simulated primary coolant condition

Wear ◽  
2013 ◽  
Vol 301 (1-2) ◽  
pp. 569-574 ◽  
Author(s):  
Young-Ho Lee ◽  
Hyung-Kyu Kim
Keyword(s):  
Nuclear Fuel ◽  
Wear Behavior ◽  
Fretting Wear ◽  
Primary Coolant ◽  
Fuel Rod

Experiments on Nonlinear Vibrations of a Nuclear Fuel Rod Supported by Spacer Grids in Air and Submerged in Water

2019 ◽  
Author(s):  
Marco Amabili ◽  
Prabakaran Balasubramanian ◽  
Giovanni Ferrari ◽  
Giulio M. Franchini ◽  
Francesco Giovanniello ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Vibration Response ◽  
Nonlinear Phenomena ◽  
Internal Resonances ◽  
Pressurized Water ◽  
Spacer Grids ◽  
Fuel Rod ◽  
Safety Margins ◽  
Fuel Assemblies ◽  
Stiffness And Damping

Abstract For safety reasons, the nuclear fuel assemblies of Pressurized Water Reactors (PWR) must be able to withstand external excitations ranging from large amplitude seismic motions of the reactor to flow-induced vibrations from the surrounding coolant water. A nuclear fuel assembly is composed of long slender tubes, most of them filled with uranium pellets, maintained in a square array by spacer grids. The spacer grids provide a nonlinear flexible boundary condition with friction and micro-impacts that complicates the nonlinear dynamics. In order to improve safety margins in the design of nuclear fuel assemblies, it is of great interest to understand the influence of the spacer grids, as it relates to the overall structural stiffness and damping properties. In particular, the evolution of the vibration amplitude with increasing excitation forces is still undetermined. In order to understand the nonlinear vibration response of a zirconium fuel rod filled with nuclear fuel pellets and supported by spacer grids, experiments were carried out in water and in air. They consisted of measuring the vibration response of the rod under a step-sine harmonic excitation at different force amplitude levels in the frequency neighborhood of the fundamental mode. If the excitation is large enough, the response of the rod displays nonlinear phenomena such as the shift of the resonant frequencies, multiple solutions with instabilities (jumps) and hysteresis, and one-to-one internal resonances. These experiments were carried out on zirconium tubes filled with axially unconstrained as well as axially blocked metallic pellets, which simulate the nuclear fuel. The zirconium tubes were tested both in air and immersed in water. The experimental data will be processed in the future by means of an identification procedure to extract the nonlinear stiffness and damping parameters of the system. An increase of the equivalent viscous damping with the excitation amplitude level is expected.

Evaluation of Fretting Wear Behavior on the Simulated Supporting Structures of a Dual-Cooled Nuclear Fuel Rod

2010 ◽  
Vol 654-656 ◽  
pp. 2564-2567
Author(s):  
Young Ho Lee ◽  
Hyung Kyu Kim
Keyword(s):  
Wear Behavior ◽  
Internal Flow ◽  
Fretting Wear ◽  
Outer Diameter ◽  
Structure Characteristics ◽  
Supporting Structure ◽  
Fuel Rod ◽  
Safety Margins ◽  
The Relationship ◽  
Supporting Structures

A dual-cooled fuel (i.e. annular fuel) has been proposed to substantially increase in power density and safety margins compared to a solid fuel in operating PWR plants. As this fuel rod has larger outer diameter than the conventional solid rod to accommodate sufficient internal flow, new supporting structure geometries should be designed and their reliabilities (i.e. vibration characteristics, fretting wear resistance, etc.) are also examined with both analytical and experimental methods. In this study, the supporting structure characteristics and fretting wear behaviors are analyzed and examined by using two kinds of simulated supporting structures that have embossing and cylindrical shapes. Their supporting structure characteristics were examined by using a specially designed test rig and their results were compared with that of analytical method. Also, fretting wear behaviors of simulated supporting structures were experimentally examined with considering the effect of contact shapes and their stiffness values. Based on the test results, the relationship between the supporting structure characteristics and their fretting wear behaviors was discussed in detail.

Measurement of Tube-to-Support Dynamic Forces in Fretting-Wear Rigs

1989 ◽  
Vol 111 (4) ◽  
pp. 385-393 ◽  
Author(s):  
N. J. Fisher ◽  
B. Ingham
Keyword(s):  
Heat Exchangers ◽  
Force Measurement ◽  
Wear Test ◽  
Dynamic Contact ◽  
Contact Forces ◽  
Fretting Wear ◽  
Test Facility ◽  
Test Results ◽  
Flow Induced Vibration ◽  
Dynamic Forces

Tubes within tube and shell heat exchangers are supported at intermediate points by plates. Flow-induced vibration can cause tubes to impact and rub against these supports, or other tubes, and can result in tube fretting-wear. The tube-to-support dynamic interaction, which consists of dynamic contact forces and tube motion, is used to relate single-span wear data to real heat exchanger configurations consisting of multi-span tube bundles. This paper describes the technique used to measure tube-to-support dynamic forces in the Chalk River Nuclear Laboratories impact fretting-wear test facility and reports test results that validate the technique. The force measurement technique is shown to provide a reasonable measure of dynamic contact forces.

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.

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.

The Predictions of the Fretting Wear between Supporting Grids and Cladding Tubes of Nuclear Fuel Rod

2006 ◽  
Vol 326-328 ◽  
pp. 1243-1246
Author(s):  
Young Chang Park ◽  
Yong Hwan Kim ◽  
Seung Jae Lee ◽  
Young Ze Lee
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Reactor ◽  
Normal Load ◽  
Fretting Wear ◽  
Wear Coefficient ◽  
Flow Induced Vibration ◽  
Stick Slip ◽  
Slip Mechanism ◽  
Fuel Rod ◽  
Wear Tests

Fretting can be defined as the oscillatory motion with very small amplitudes, which usually occur between two contacting surfaces. Fretting wear is the removal of material from contacting surfaces through fretting action. This 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 performed with two types of cladding tubes and three types of supporting grids in water. Fretting wear tests were done using various applied loads. 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. Depending on various normal load, tube materials, and supporting grid shapes, distinctively different wear scar of fretting and stick-slip mechanism can occur.

scholarly journals Fretting Wear Behavior of Three Kinds of Rubbers under Sphere-On-Flat Contact

Materials ◽  
2021 ◽  
Vol 14 (9) ◽  
pp. 2153
Author(s):  
Tengfei Zhang ◽  
Jie Su ◽  
Yuanjie Shu ◽  
Fei Shen ◽  
Liaoliang Ke
Keyword(s):  
Normal Force ◽  
Wear Behavior ◽  
Displacement Amplitude ◽  
Fretting Wear ◽  
Butadiene Rubber ◽  
Force Frequency ◽  
Sealing Performance ◽  
Rubber Surface ◽  
Sealing Materials ◽  
Flat Contact

Rubbers are widely used in various fields as the important sealing materials, such as window seal, door seal, valve, pump seal, etc. The fretting wear behavior of rubbers has an important effect on their sealing performance. This paper presents an experimental study on the fretting wear behavior of rubbers against the steel ball under air conditions (room temperature at 20 ± 2 °C and humidity at 40%). Three kinds of rubbers, including EPDM (ethylene propylene diene monomer), FPM (fluororubber), and NBR (nitrile–butadiene rubber), are considered in experiments. The sphere-on-flat contact pattern is used as the contact model. The influences of the displacement amplitude, normal force, frequency, and rubber hardness on the fretting wear behavior are discussed in detail. White light profiler and scanning electron microscope (SEM) are used to analyze the wear mechanism of the rubber surface. The fretting wear performances of three rubbers are compared by considering the effect of the displacement amplitude, normal force, frequency, and rubber hardness. The results show that NBR has the most stable friction coefficient and the best wear resistance among the three rubbers.

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