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Lattice physics codes are primarily used to generate cross-section data for nodal codes. In this work the methodology of homogenized constant generation was applied to a small Pressurized Water Reactor (PWR) core, using the deterministic code Helios and the Monte Carlo code Serpent. Subsequently, a 3D analysis of the PWR core was performed with the nodal diffusion code PARCS using the two-group cross section data sets generated by Helios and Serpent. Moreover, a full 3D model of the PWR core was developed using Serpent in order to obtain a reference solution. Several parameters, such as k eff , axial and radial power, fission and capture rates were compared and found to be in good agreement. [ABSTRACT FROM AUTHOR]

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]

• Experimental study on vibrations of reduced-scale PWR fuel bundles. • Large amplitude vibrations in non-linear regime. • Evidence of damping increase with vibration amplitude. • Spacer grids as boundary conditions with nonlinear stiffness and friction. • Tests in presence of water flowing at typical coolant flow velocity. The core of Pressurized Water Reactors (PWR) is composed by nuclear fuel assemblies, bundles of slender fuel rods containing uranium pellets, kept in position by spacer grids equipped with elastic elements for retention, and exposed to an axial flow of coolant. Fuel assemblies are subjected to fluid-induced vibrations causing fretting at the interface with spacer grids; events such as earthquakes may also constitute an external excitation resulting in large-amplitude vibrations of the fuel rods. The relationship between the excitation amplitude and the damping during large-amplitude forced vibrations of nuclear fuel assemblies in the presence of flowing fluid is not fully understood. The present experimental study investigated the vibrations of a 3 × 3 tube assembly composed by eight regularly spaced fuel rods installed around a guide tube and supported by spacer grids; tests were performed in still and flowing water. In addition to experimental modal analysis under small random excitation, stepped-sine experiments at different levels of harmonic excitation that caused large-amplitude vibrations were performed. The equivalent viscous damping ratios of the fundamental model at different excitation levels were extracted by fitting the results with a single degree-of-freedom model. The vibrations of the fuel assembly were strongly influenced by the vibrational behavior of the single rods, which constitute coupled oscillators. An increase of damping with the excitation amplitude was observed both in still and in flowing water and acted in the direction of structural safety. The water flow did not cause instabilities in the operational range; instead, the increment of flow speed increased the damping ratios in the linear (small-amplitude vibrations) and nonlinear (large-amplitude vibrations) regime. [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]