Effect of Clamping Failure on Flow Induced Vibration and Fretting Wear of Fuel Rods

2018 ◽  
Author(s):  
Qi Huan-huan ◽  
Feng Zhi-peng ◽  
Xiong Fu-rui ◽  
Jiang Nai-bin ◽  
Huang Qian ◽  
...  
Keyword(s):  
Natural Frequency ◽  
Lateral Flow ◽  
Fretting Wear ◽  
Leaf Spring ◽  
Wear Depth ◽  
Flow Induced Vibration ◽  
Vibration Displacement ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Rod

Fuel rods are subjected to both axial and lateral flow in the reactor core. In this study, we present a study on the flow induced vibration (FIV) and fretting wear of fuel rod with failed clamping at grids. First, 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. Next, by combining the correlation 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 clamping of fuel rod at various grids may fail due to inaccurate manufacturing, fuel transportation and in-core irradiation. The absence of 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 response amplitude of turbulent excitation at the bottom and top of the fuel rod is 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 over the rod wear depth depicts basically the same trend as on the maximum FIV amplitude. In addition to the FIV amplitude, wear depth is also related to rod natural frequency. By examining the multiplication of amplitude and frequency at the top and bottom grid, we found that the dimple failure has greater impact at the top grid.

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.

Fretting Wear of the Zr Fuel Rod Cladding

2014 ◽  
Author(s):  
Ladislav Pecinka ◽  
Jaroslav Svoboda ◽  
Vladimír Zeman
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactors ◽  
Fretting Wear ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Fuel Assemblies ◽  
Flow Induced Vibrations ◽  
Water Reactors ◽  
Fretting Damage

Fretting wear is a particular type of wear that is expected to occur in fuel assemblies of pressurized water cooled nuclear reactors. Fretting damage of fuel rods may cause Nuclear Power Plant (NPP) operations problems and are very expensive to repair. As utilities and fuel vendors adopt higher utilization of uranium and improved thermal margins plants, burned fuel rods will be loaded at core the periphery as part of the margin mechanisms. Pressurized Water Reactors (PWRs) have experienced fuel rods fretting wear failures due to flow induced vibrations. This study describes basic results of theoretical analysis and describes experiments to predict thinning of the Zr cladding wall thickness performed.

Research on the Effect of Grid Clamping Failure on Flow Elastic Stability and Vortex Shedding of Fuel Rod

2018 ◽  
Author(s):  
Qi Huan-huan ◽  
Feng Zhi-peng ◽  
Jiang Nai-bin ◽  
Huang Qian ◽  
Huang Xuan
Keyword(s):  
Natural Frequency ◽  
Vortex Shedding ◽  
Vibration Analysis ◽  
Natural Frequencies ◽  
Elastic Stability ◽  
Mode Shapes ◽  
Flow Induced Vibration ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Fuel Assemblies

Flow elastic stability and vortex shedding were two important mechanisms of the flow induced vibration analysis. Due to the influence of manufacturing process, transportation and irradiation, the clamping action of grid on fuel rods may be invalid. Taking one fuel assemblies as an example, the effects of clamping failure on the natural frequencies, mode shapes, flow elastic stability and vortex shedding were studied. The results show that the effect of the rigid convex support failure on the natural frequency was directly related to the mode shape. The effect of the grid rigid convex failure near the nodes with larger amplitude on the natural frequency was obvious. The velocity of flow at the top and bottom of the fuel rods were larger and the size was comparable, this induced that the rigid convex failure of the top and bottom grids had a significant effect on the ratio of flow elastic stability and the vortex shedding.

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.

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.

Fretting Wear Between Supporting Grids and Cladding Tubes of Nuclear Fuel Rod

2006 ◽  
Author(s):  
Young-Chang Park ◽  
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 ◽  
Fuel Rod ◽  
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°C, 50°C and 80°C were tested with the applied loads from 5N up to 25N and the relative amplitude of 200μm. The worn surfaces were observed by SEM, EDX analysis and 2D surface profiler. As the water temperature increased, the wear volume was decreased, but oxide layer was increased on the worn surface. The abrasive wear mechanism was observed at water temperature of 20°C and adhesive wear mechanism occurred at water temperature of 50°C, 80°C. As the water temperature increased, surface micro-hardness was decreased, but 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.

Monte Carlo Simulation of Fretting Wear Performance of Fuel Rods

2005 ◽  
Author(s):  
Pablo R. Rubiolo
Keyword(s):  
Monte Carlo ◽  
Fretting Wear ◽  
Fuel Rods ◽  
Wear Performance ◽  
Wear Rates ◽  
Fuel Rod ◽  
Cell Clearance ◽  
Vibration Model ◽  
Support Conditions ◽  
Wear Damage

The effect of the diverse parameters affecting the fretting-wear performance of nuclear fuel rods is investigated by performing Monte Carlo simulations with a fuel rod vibration model. The study is focused on the analysis of the effect of the grid parameters, including the cell clearance and the grid/support misalignments, on the support preload forces distribution, the rod dynamic response and the overall wear damage. In the present approach, the fuel rod and grids are modeled as a beam constrained at a finite number of axial positions and a non-linear vibration model is used to predict the rod motion and the wear rates. The results of the analysis suggest that an important fraction of the variability of the assembly wear damage distribution can be explained by the local variations of the rod-support conditions.

Multiphysics Modeling of Pressurized Water Reactor Fuel Performance

2018 ◽  
Vol 4 (3) ◽  
Author(s):  
Wang Zhu ◽  
Zhang Chungyu ◽  
Yuan Cenxi
Keyword(s):  
Three Dimensional ◽  
Fission Products ◽  
Axial Direction ◽  
General Purpose ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Nuclear Fuel Rods ◽  
Reactivity Insertion

Nuclear fuel rods operate under complex radioactive, thermal, and mechanical conditions. Nowadays, fuel rod codes usually make great simplifications on analyzing the multiphysics behavior of fuel rods. The present study develops a three-dimensional (3D) module within the framework of a general-purpose finite element solver, i.e., abaqus, for modeling the major physics of the fuel rods. A typical fuel rod, subjected to stable operations and transient conditions, is modeled. The results show that the burnup levels have an important influence on the thermomechanical behavior of fuel rods. The swelling of fission products causes a dramatically increasing strain of pellets. The variation of the stress and the radial displacement of the cladding along the axial direction can be reasonably predicted. It is shown that a quick power ramp or a reactivity insertion accident can induce high tensile stress in the outer regime of the pellet and may cause further fragmentation to the pellets. Fission products migration effects and differential thermal expansion become more severe if there are flaws or imperfections on the pellet.

Flow-Induced Grid-to-Rod Fretting Test and Simulation for NFI PWR Fuel Assembly

2013 ◽  
Author(s):  
Shota Okui ◽  
Yuichiro Kubo ◽  
Shumpei Kakinoki ◽  
Roger Y. Lu ◽  
Zeses Karoutas ◽  
...  
Keyword(s):  
Numerical Simulation ◽  
Fuel Assembly ◽  
Wear Test ◽  
Flow Induced Vibration ◽  
Fuel Rods ◽  
Spacer Grids ◽  
Fuel Rod ◽  
Non Linear ◽  
Computational Fluid Dynamics Cfd

A long-term flow-induced vibration and wear test was performed for a full-scale 17×17 PWR fuel mockup, and the test results were compared with numerical simulations. The flow-induced vibration on a fuel assembly or fuel rods may cause Grid-to-Rod Fretting (GTRF) and result in the leakage of fuel rods in PWRs. GTRF involves non-linear vibration of a fuel rod due to the excitation force induced by coolant flow around a fuel rod. So, the numerical simulation is performed by VITRAN (Vibration Transient Analysis Non-linear) and Computational Fluid Dynamics (CFD). VITRAN code was developed by Westinghouse to simulate fuel rod flow induced vibration and GTRF. In this paper, it was confirmed that the code can reproduce GTRF wear for NFI fuel assembly. CFD calculation is performed to obtain the axial and lateral flow velocity around the fuel rods, reflecting detailed geometries of fuel assembly components like bottom nozzle, spacer grids. The numerical simulation reasonably reproduced the vibration and wear test for NFI fuel assembly.

Fretting Wear Behavior of APMT Steel at 350°C for Reactor Fuel Cladding Application

MRS Advances ◽  
2016 ◽  
Vol 1 (35) ◽  
pp. 2495-2500
Author(s):  
Thomas Winter ◽  
James Huggins ◽  
Richard Neu ◽  
Preet Singh ◽  
Chaitanya S. Deo
Keyword(s):  
Nuclear Reactor ◽  
Wear Behavior ◽  
Sem Analysis ◽  
Fretting Wear ◽  
Pressurized Water Reactor ◽  
Fuel Cladding ◽  
Reactor Accident ◽  
Pressurized Water ◽  
Water Reactor ◽  
Fuel Rod

ABSTRACTIn support of a recent surge in research to develop an accident tolerant reactor, accident tolerant fuels and cladding candidates are being investigated. Relative motion between the fuel rods and fuel assembly spacer grids can lead to excessive fuel rod wear and, in some cases, to fuel rod failure. Based on industry data, grid-to-rod-fretting (GTRF) has been the number one cause of fuel failures within the U.S. pressurized water reactor (PWR) fleet, accounting for more than 70% of all PWR leaking fuel assemblies. APMT, an Fe-Cr-Al steel alloy, is being examined for the I2S-LWR project as a possible alternative to conventional fuel cladding in a nuclear reactor due to its favorable performance under LOCA conditions. Tests were performed to examine the reliability of the cladding candidate under simulated fretting conditions of a pressurized water reactor (PWR). The contact is simulated with a rectangular and a cylindrical specimen over a line contact area. A combination of SEM analysis and wear & work rate calculations are performed on the samples to determine their performance and wear under fretting. While APMT can perform favorably in loss of coolant accident scenarios, it also needs to perform well when compared to Zircaloy-4 with respect to fretting wear.

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