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.

Prediction of the Nuclear Fuel Rod Abrasion in the Probability Sense

2016 ◽  
Vol 821 ◽  
pp. 317-324
Author(s):  
Vladimír Zeman ◽  
Zdeněk Hlaváč
Keyword(s):  
Fuel Assembly ◽  
Nuclear Fuel ◽  
Fretting Wear ◽  
Primary Circuit ◽  
Grid Cells ◽  
Operational Parameters ◽  
Spacer Grid ◽  
Mean Values ◽  
Fuel Rod ◽  
Lower Limits

The paper deals with the upper and lower limits estimation of the friction work and fretting wear in the contact of nuclear fuel rods with fuel assembly (FA) spacer grid cells. The friction work is deciding factor for the prediction of the fuel rod cladding abrasion caused by FA vibration. Design and operational parameters of the FA components are understood as random variables defined by mean values and standard deviations. The gradient and three sigma criterion approach is applied to the calculation of the upper and lower limits of the friction work and fretting wear in particular contact surfaces between the fuel rod cladding and some of spacer grid cells. The fuel assembly vibration is excited by pressure pulsations of the cooling liquid generated by main circulation pumps in the coolant loops of the NPP primary circuit. The method is applied for hexagonal type nuclear fuel assembly in the VVER type reactors.

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.

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.

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

Identification of Non-Linear Vibration Parameters of a Nuclear Fuel Rod

2018 ◽  
Author(s):  
Marco Amabili ◽  
Prabakaran Balasubramanian ◽  
Giovanni Ferrari ◽  
Stanislas Le Guisquet ◽  
Kostas Karazis ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Harmonic Balance Method ◽  
Resonance Phenomenon ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Fuel Pellets ◽  
Safety Margins ◽  
Fuel Assemblies ◽  
Non Linear

In Pressurized Water Reactors (PWR), fuel assemblies are composed of fuel rods, long slender tubes filled with uranium pellets, bundled together using spacer grids. These structures are subjected to fluid-structure interactions, due to the flowing coolant surrounding the fuel assemblies inside the core, coupled with large-amplitude vibrations in case of external seismic excitation. Therefore, understanding the non-linear response of the structure and, particularly, its dissipation, is of paramount importance for the choice of safety margins. To model the nonlinear dynamic response of fuel rods, the identification of nonlinear stiffness and damping parameters is required. The case of a single fuel rod with clamped-clamped boundary conditions was investigated by applying harmonic excitation at various force levels. Different configurations were implemented testing the fuel rod in air and in still water; the effect of metal pellets simulating nuclear fuel pellets inside the rods was also recorded. Non-linear parameters were extracted from some of the experimental response curves by means of a numerical tool based on the harmonic balance method. The axisymmetric geometry of fuel rods resulted in the presence of a one-to-one internal resonance phenomenon, which has to be taken into account modifying accordingly the numerical identification tool. The internal motion of fuel pellets is a cause of friction and impacts, complicating further the linear and non-linear dynamic behavior of the system. An increase of the equivalent viscous-based modal damping with excitation amplitude is often shown during geometrically non-linear vibrations, thus confirming previous experimental findings in the literature.

Role of ZrO2 oxide layer on the fretting wear resistance of a nuclear fuel rod

2020 ◽  
Vol 145 ◽  
pp. 106146 ◽  
Author(s):  
Young-Ho Lee ◽  
Il-Hyun Kim ◽  
Hyung-Kyu Kim ◽  
Hyun-Gil Kim
Keyword(s):  
Wear Resistance ◽  
Oxide Layer ◽  
Nuclear Fuel ◽  
Fretting Wear ◽  
Fuel Rod

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.

Experimental Study on the Wear Behavior of Nuclear Fuel Rod in High Temperature Water Environment

2005 ◽  
Vol 297-300 ◽  
pp. 1395-1400
Author(s):  
Young Ho Lee ◽  
Hyung Kyu Kim ◽  
Youn Ho Jung
Keyword(s):  
High Temperature ◽  
Nuclear Fuel ◽  
Wear Behavior ◽  
Water Environment ◽  
Wear Particle ◽  
Fuel Rod ◽  
High Temperature Water ◽  
Particle Layer ◽  
Increasing Temperature ◽  
Temperature Water

In this study, the variation of spring characteristics with increasing temperature was examined and the effect of their variations on the wear behavior of a nuclear fuel rod in both room and high temperature (300°C) water conditions was evaluated. From the results of the load-displacement tests, the spring stiffness was remarkably varied with increasing temperature. The results of the wear tests indicated that the wear damages are decreased at high temperature water when compared with the room temperature result. These results indicated that the removal mechanisms of wear debris at high temperature water are dependent on not only the formation of the wear particle layer but also on the changed contact conditions such as the contact length or area due to the stiffness drops.

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