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.

Identification of Non-Linear Parameters of a Nuclear Fuel Rod

2017 ◽  
Author(s):  
Giovanni Ferrari ◽  
Stanislas Le Guisquet ◽  
Prabakaran Balasubramanian ◽  
Marco Amabili ◽  
Brian Painter ◽  
...  
Keyword(s):  
Circular Cylindrical Shell ◽  
Harmonic Balance Method ◽  
Excitation Amplitude ◽  
Resonance Phenomenon ◽  
Response Curves ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Safety Margins ◽  
Fuel Assemblies

In Pressurized Water Reactors (PWR), fuel assemblies are made up 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 nonlinear response of the structure, and, particularly, its dissipation, is of paramount importance for the choice of safety margins, in the design of fuel assemblies, to ensure their functionality and safety in the worst external condition scenarios. To model the nonlinear dynamic response of fuel rods, the identification of the nonlinear stiffness and damping parameters is required. A tool based on harmonic balance method was developed to identify these parameters from the experimentally obtained force-response curves, considering one-to-one internal resonance phenomenon present in axisymmetric structures such as cylindrical tubes and shells. To validate the tool, it was applied to the reference case of circular cylindrical shell filled with water, which revealed an increase of damping with the excitation amplitude. In the following paper, the more realistic case of a single fuel rod with clamped-clamped boundary condition was investigated by applying harmonic excitation at various force levels. The nonlinear parameters including damping were extracted from experimental results by means of the adapted tool. An increase in damping with excitation amplitude has been shown according to earlier studies.

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.

Rod Bow Ultrasonic Measurements Possibilities

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
Marcin Kopeć ◽  
Martina Malá
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Reactors ◽  
Research Centre ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Nuclear Fuel Assembly ◽  
Fuel Assemblies ◽  
Ultrasonic Measurements ◽  
History Of ◽  
Fuel Particles

The ultrasonic (UT) measurements have a long history of utilization in the industry, also in the nuclear field. As the UT transducers are developing with the technology in their accuracy and radiation resistance, they could serve as a reliable tool for measurements of small but sensitive changes for the nuclear fuel assembly (FA) internals as the fuel rods are. The fuel rod bow is a phenomenon that may bring advanced problems as neglected or overseen. The quantification of this issue state and its probable progress may help to prevent the safety-related problems of nuclear reactors to occur—the excessive rod bow could, in the worst scenario, result in cladding disruption and then the release of actinides or even fuel particles to the coolant medium. Research Centre Rez has developed a tool, which could serve as a complementary system for standard postirradiation inspection programs for nuclear fuel assemblies. The system works in a contactless mode and reveals a 0.1 mm precision of measurements in both parallel (toward the probe) and perpendicular (sideways against the probe) directions.

Non-Linear Damping Identification in Nuclear Systems Under External Excitation

2015 ◽  
Author(s):  
Joachim Delannoy ◽  
Marco Amabili ◽  
Brett Matthews ◽  
Brian Painter ◽  
Kostas Karazis
Keyword(s):  
Vibration Amplitude ◽  
Harmonic Balance Method ◽  
Harmonic Excitation ◽  
Metal Plate ◽  
Nuclear Industry ◽  
External Excitation ◽  
Pressurized Water ◽  
Vibration Data ◽  
Fuel Assemblies ◽  
Non Linear

In Pressurized Water Reactors (PWR) the fluid-structure interaction between the coolant and fuel assemblies is an important phenomenon that is directly associated to safety and performance issues for the nuclear industry. Fuel assemblies are formed by bundling fuel rods, long slender pressurized tubes containing uranium pellets, with hollow guide and instrumentation tubes using a number of spacer grids for support. In order to correctly simulate the response of a fuel assembly under external excitation, an investigation of the system damping is necessary. Recent study shows that the damping may grow significantly with the vibration amplitude, increasing the safety factor. In order to address this issue, the present study has developed a tool to identify the vibration characteristics of non-linear mechanical systems from experimental forced vibration data obtained at different excitation levels. In particular, this project focuses on the damping identification. A parameter identification methodology for non-linear systems, based on harmonic balance method, is used to identify the damping governing the motion of such systems. The tool has been developed by using hardening non-linear responses of two sandwich panel and a metal plate subjected to external harmonic excitation. The method has been validated by comparison with the damping identified by the full non-linear model of the two sandwich panels. In the three cases, an increase of damping with the vibration amplitude is found and discussed.

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.

Enhancement of Heat Transfer Performance in Nuclear Fuel Rod Using Nanofluids and Surface Roughness Technique

2015 ◽  
Author(s):  
Kang Liu ◽  
Titan C. Paul ◽  
Leo A. Carrilho ◽  
Jamil A. Khan
Keyword(s):  
Heat Transfer ◽  
Surface Roughness ◽  
Heat Transfer Coefficient ◽  
Transfer Coefficient ◽  
Nuclear Fuel ◽  
Three Dimensional ◽  
Pressurized Water ◽  
Bulk Fluid ◽  
Fuel Rod ◽  
Dimensional Surface

The experimental investigations were carried out of a pressurized water nuclear reactor (PWR) with enhanced surface using different concentration (0.5 and 2.0 vol%) of ZnO/DI-water based nanofluids as a coolant. The experimental setup consisted of a flow loop with a nuclear fuel rod section that was heated by electrical current. The fuel rod surfaces were termed as two-dimensional surface roughness (square transverse ribbed surface) and three-dimensional surface roughness (diamond shaped blocks). The variation in temperature of nuclear fuel rod was measured along the length of a specified section. Heat transfer coefficient was calculated by measuring heat flux and temperature differences between surface and bulk fluid. The experimental results of nanofluids were compared with the coolant as a DI-water data. The maximum heat transfer coefficient enhancement was achieved 33% at Re = 1.15 × 105 for fuel rod with three-dimensional surface roughness using 2.0 vol% nanofluids compared to DI-water.

A Comparison of Fast Thorium Breeder Reactor Designs With Oxide and Metallic Fuels

2008 ◽  
Author(s):  
V. Jagannathan ◽  
Usha Pal ◽  
R. Karthikeyan ◽  
Devesh Raj
Keyword(s):  
Unit Volume ◽  
Core Region ◽  
Metallic Fuel ◽  
Fuel Rods ◽  
Fuel Rod ◽  
The Core ◽  
Fuel Assemblies ◽  
Depletion Rate ◽  
Metallic Fuels ◽  
Core Characteristics

Loading of seedless thoria rods in internal blanket regions and using them later as part of seeded fuel assemblies is the central theme of the thorium breeder reactor (ATBR) concept [1]. The fast reactors presently consider seedless blanket region surrounding the seeded core region. This results in slower fissile production rate in comparison to fissile depletion rate per unit volume. The overall breeding is achieved mainly by employing blanket core with more than double the volume of seeded core. The blanket fuel is discharged with fissile content of ∼30g/kg, which is much less than the asymptotic maximum possible fissile content of 100g/kg. This is due to smaller coolant flow provided for in the blanket regions. In a newly proposed fast thorium breeder reactor (FTBR) [2], the blanket region is brought in and distributed through out the core. By this the fissile depletion and production rates per unit volume become comparable. The core considered simultaneous breeding from both fertile thoria and depleted uranium and hence the concept can be called as fast twin breeder reactor as well. Sodium is used as coolant. The blanket fuel rods achieve nearly 80% of the seed fuel rod burnup and also contain nearly the maximum possible fissile content at the time of discharge. In this paper a comparison of FTBR core characteristics with oxide and metallic fuel are compared.

scholarly journals Experimental Approaches for Manufacturing of Simulated Cladding and Simulated Fuel Rod for Mechanical Decladder

2020 ◽  
Vol 2020 ◽  
pp. 1-12
Author(s):  
Young-Hwan Kim ◽  
Yung-Zun Cho ◽  
Jin-Mok Hur
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Energy Research ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Compressive Stresses ◽  
Simulated Fuel ◽  
Ceramic Fuel ◽  
Experimental Approaches

We are developing a practical-scale mechanical decladder that can slit nuclear spent fuel rod-cuts (hulls + pellets) on the order of several tens of kgf of heavy metal/batch to supply UO2 pellets to a voloxidation process. The mechanical decladder is used for separating and recovering nuclear fuel material from the cladding tube by horizontally slitting the cladding tube of a fuel rod. The Korea Atomic Energy Research Institute (KAERI) is improving the performance of the mechanical decladder to increase the recovery rate of pellets from spent fuel rods. However, because actual nuclear spent fuel is dangerously toxic, we need to develop simulated spent fuel rods for continuous experiments with mechanical decladders. We describe procedures to develop both simulated cladding tubes and simulated fuel rod (with physical properties similar to those of spent nuclear fuel). Performance tests were carried out to evaluate the decladding ability of the mechanical decladder using two types of simulated fuel (simulated tube + brass pellets and zircaloy-4 tube + simulated ceramic fuel rod). The simulated tube was developed for analyzing the slitting characteristics of the cross section of the spent fuel cladding tube. Simulated ceramic fuel rod (with mechanical properties similar to the pellets of actual PWR spent fuel) was produced to ensure that the mechanical decladder could slit real PWR spent fuel. We used castable powder pellets that simulate the compressive stress of the real spent UO2 pellet. The production criteria for simulated pellets with compressive stresses similar to those of actual spent fuel were determined, and the castables were inserted into zircaloy-4 tubes and sintered to produce the simulated fuel rod. To investigate the slitting characteristics of the simulated ceramic fuel rod, a verification experiment was performed using a mechanical decladder.

Numerical Analysis of Transient Temperature Distribution in a Partially Cooled Nuclear Fuel Rod

2015 ◽  
Author(s):  
Sandeep Patil ◽  
Siddarth Chintamani ◽  
Rajeev Kumar ◽  
Ratan Kumar ◽  
Brian H. Dennis
Keyword(s):  
Temperature Distribution ◽  
Numerical Analysis ◽  
Nuclear Fuel ◽  
Learning Curves ◽  
Coolant Flow ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Coolant Water ◽  
Nuclear Fuel Rods ◽  
Temperature Levels

Critical safety studies of a nuclear power plants are often associated with inadequate and improper cooling of the reactor core or the spent fuel rods. Coolant flow over the hot nuclear fuel rods often gets stalled during major accidents resulting in high temperature levels. These elevated temperature levels can potentially melt the fuel rod material and cause the release of radioactive gases. Research activities, both numerical and experimental in nature to explore these rare but potentially catastrophic possibilities have resulted in sophisticated numerical codes capable of simulating the various post-accident scenarios. These codes, although reasonably accurate and reliable have steep learning curves and are not often very user-friendly. A fast and accurate prediction of the critical temperature conditions using popular commercially available software packages is the subject of current study. Results from this parametric study of temperature distribution over a partially cooled fuel rod carried out using ANSYS as the numerical analysis tool is reported. Nuclear fuel rods being inadequately cooled inside a stagnant pool of coolant water in an accident scenario resulting in disrupted coolant flow has been simulated. This situation can arise within the reactor (design-basis accidents) or in the waste-fuel storage (as faced in Fukushima). In these situations, the fuel rod is often left partially immersed in the coolant water resulting in immersed portion of the rod cooled by water and the exposed portion cooled by air leading to non-uniform and improper cooling of the system. Realistic dimensions and materials as in commercial nuclear fuel rod have been used in the study. Taking advantage of the symmetry, an axisymmetric radial plane sliced longitudinally has been analyzed. Variations in the tangential direction have been neglected. The heat transfer problem uses homogeneous convective boundary conditions and assumes temperature dependent thermal conductivity. The parameters varied are the coolant level and the heat generation rate inside the fuel rod. A macro to automatically capture the transients in the temperatures was written in ANSYS (a finite element package). The governing energy equations were implicitly solved using finite volume scheme in MATLAB. ANSYS results are in close agreement with those obtained using MATLAB. The centerline temperature of the fuel rod shows a sharp rise below a certain coolant level.

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.

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