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Nuclear fuel rods prevent the loss of nuclear material to the refrigerant fluid, acting as a first safety barrier. Damage by grid-to-rod fretting remains a significant cause of nuclear fuel element failure in Pressurized Water Reactors (PWR). In this paper fretting-induced friction at the contact interface between springs and samples of nuclear fuel rods of Zr-1.0Nb-1.0Sn-0.1Fe alloy was evaluated. The effect on the fretting resistance of heat treatments applied to the samples of Zr-1.0Nb-1.0Sn-0.1Fe fuel rods was also investigated. It was observed that the increase in the heat treatment temperature decreases the mass loss, indicating a mild fretting wear regime. [Display omitted] • The nuclear fuel rods prevent the loss of nuclear material to the refrigerant fluid. • Damage by grid-to-rod fretting remains a significant cause of nuclear fuel failure. • Heat treatment promoted an increase in the Zr–1Nb–1Sn-0.1Fe fretting resistance. [ABSTRACT FROM AUTHOR]

[Display omitted] • Full 3D-tomography of boiling annular flow. • Closed flow-loop with heat transfer and chloroform as working fluid. • Various steady-state conditions including dryout measured. • Validation of film thickness reconstruction method. • Complex and non-linear behavior of liquid film found. For the validation of advanced two-fluid CFD models at boiling water reactor (BWR) conditions, it is necessary to have good experimental data challenging for the code by reflecting relevant flow phenomena. The present experimental study aims to provide detailed film thickness distributions in a BWR-like subchannel geometry, including spacer grids and heat transfer. This is achieved using chloroform as a working fluid, indirectly heated by hot water channels simulating BWR fuel rods. X-ray equipment was used to produce tomographic 3D-data of time-averaged attenuation coefficients which essentially represent the liquid holdup. The conversion to film thickness was done by integrating the attenuation coefficient along lines perpendicular to the rod wall and using an effective chloroform attenuation coefficient determined by numerical simulation of the experimental setup, including X-ray sensor behavior. The simulation method was validated by using two different sensors types. Measurements were performed over the whole cross-section of the channel and an axial length of 11 cm. The experiments cover a range of liquid flow rates in the pre- and post-dryout regime. [ABSTRACT FROM AUTHOR]

A mechanical apparatus was developed to investigate the grid-to-rod fretting on nuclear assembly of Zr–1Nb–1Sn-0.1Fe alloy. Test apparatus allowed the performing of fretting tests on fuel rods at a frequency to 30 Hz, which was evaluated as the first natural frequency of a fuel rod. In addition, strain-gauges was employed to evaluate the misalignment effect on the mechanical behavior of the testing material. The surface damage intensity decreased from the central part, to the extremity of wear areas. The variation of wear areas revealed that the misalignment promoted premature wear because the intensification of fretting wear occurred by the appearance of complementary efforts. The confocal method suggested that wear residues under dry conditions remained adhered to the fretting surface. • The greater the rod misalignment over time, the greater the imbalance of its structural behavior to wear. • Areas of wear presented greater superficial damage in its central part evolving in a lower intensity to the extremities. • The confocal method suggests that the wear residues in the dry condition remained adhered to the worn surface. • Misalignment promotes an intensification of the fretting wear by the appearance of complementary efforts. [ABSTRACT FROM AUTHOR]

• Experimental study on vibrations of zirconium alloy nuclear fuel rods. • Large amplitude vibrations in non-linear regime. • Evidence of damping increase with vibration amplitude. • Presence of pellets, tests in air and submerged in water. • Spacer grids as boundary conditions with nonlinear stiffness and friction. The internal components of Pressurized Water Reactors (PWRs), particularly nuclear fuel assemblies, must be able to withstand Flow Induced Vibrations (FIVs) during operating conditions and during extreme accident conditions, such as earthquakes. Nuclear fuel assemblies are composed of long slender tubes, filled with uranium pellets that are bundled together by periodic support provided by spacer grids. Spacer grids are square structures used to increase thermal mixing in the core and provide support to the fuel rods and guide tubes allowing for the installation and removal of nuclear fuel rods. Nevertheless, spacer grids constitute a nonlinear flexible boundary condition experiencing friction forces and impacts complicating the dynamics of the fuel rod-spacer grid system. In order to improve safety margins in the design of nuclear fuel assemblies, it is of great interest to understand the nonlinear behavior at the interface of the spacer grids with the fuel rods, investigating the complexity due to the nonlinear evolution of the system stiffness and damping properties. In particular, the effect of the constant evolution of the vibration amplitude as a function of the change of the excitation forces on the dynamics of the fuel rod response is still undetermined. Experiments were carried out in quiescent water and in air to understand the nonlinear vibration response of a single zirconium fuel rod supported by spacer grids. The vibration response under a step-sine harmonic excitation at different force amplitude levels in the frequency neighborhood of the fundamental mode was measured. The response of the rod displayed nonlinear phenomena such as the shift of the resonant frequencies, multiple solutions with some instabilities (jumps) and hysteresis, and a weak one-to-one internal resonance. Tests were performed on an empty rod and on a rod filled with tungsten pellets representative of nuclear fuel. The pellets were let free to move and were subsequently blocked axially to reproduce the effect of the beginning-of-life constraint in operational nuclear plants. The experimental data were processed by means of a simplified identification procedure to extract the damping parameters of the vibrating system. The equivalent viscous damping is found to increase and to be a function of the level of excitation and of the peak vibration amplitude. [ABSTRACT FROM AUTHOR]