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For the modeling of the interfacial structure characteristics, an experiment of vertical adiabatic air-water flow has been conducted under an atmosphere pressure condition. The experimental facility is composed of 5 × 5 rods arranged on a square pitch in a square casing with 9.5 mm outside rod diameter and 12.6 mm pitch, simulating a prototypic rod bundle in nuclear reactors. The miniaturized four-sensor conductivity probe has been developed to allow for the experimental measurement in the small flow channel. Extensive data are acquired for one-dimensional flow parameters including axial development of void fraction, interfacial area concentration and gas velocity. The effect of a prototypic spacer grids with mixing vanes on the flow structure has been discussed at various flow conditions based on the experimental data. Two competing effects, “swirling effect” and “bubble breaker effect”, are identified. The effect vanishes out within a short distance from the space grid. A drift-flux correlation is developed for predicting void fraction of adiabatic bubbly two-phase flow in a vertical rod bundle. Existing interfacial area correlations are tested and Hibiki-Ishii correlation (2002b) is recommended as most accurate correlation to predict the interfacial area concentration. It is expected that the newly obtained data in the 5 × 5 rod bundle is useful for developing the interfacial area transport equation and benchmarking a computational code. [ABSTRACT FROM AUTHOR]

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]

[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]

• We conducted a heat transfer experiment of SC water flowing upward in 2 × 2 rod bundles. • We used 3 different methods to calculate the HTC of sub-channels of rod bundles. • Spacer effects on heat transfer was observed and CFD simulation of spacer effect was also investigated. • HTCs in 2 × 2 rod bundles were compared with the annuli and circular tube of the same equivalent hydraulic diameter. This paper presents the investigation of heat transfer of supercritical water flowing upward in 2 × 2 rod bundles, both experimentally and numerically. The experiment conditions are as follows: pressures from 22.9 MPa to 26.1 MPa, mass fluxes ranging from 450 kg/m2/s to 1500 kg/m2/s and heat fluxes ranging from 0.4 MW/m2 to 1.5 MW/m2. Four heated-rods of 8 mm outer diameter formed 2 set of experiment sections with pitch to diameter ratio (P/D ratio) 1.18 and 1.3. Heat transfer coefficients between the heated rod and fluid were conducted by 3 different methods and the effect of spacer grids was clearly observed. The experiment results show that heat transfer to supercritical water in rod bundles is strongly affected by flow channels and spacer grids. Moreover, a large database of heat transfer in rod bundles was carried out and HTC were compared with tube and annuli with the equivalent hydraulic diameter. The phenomenon of heat transfer difference in rod bundle circumferential direction and enhancement heat transfer in downstream of spacer grids were observed and investigated. In addition, Computational Fluid Dynamics (CFD) method was applied to give a deep insight of velocity profile, spacer effect and heat transfer enhancement in this research. [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]

The vibration of rods is of prime importance to estimate their wearing due to the frictions with grids' elements. To predict this wearing, the pressure fluctuations exerted on the rods need to be identified and quantified. Pressure fluctuations are explored using an instrumented rod centred within a 5 × 5 rods bundle. The profiles of mean velocity and velocity fluctuations are explored around the instrumented rod using LDV. Two configurations of spacer-grids are used: with and without mixing vanes respectively noted WMV and NMV. The position of the measurement point is rotated over 360° and its distance from the grids is varied from 0.5D h to 20D h. The results are detailed for the Reynolds number 66000. The results show the non-homogeneity of the azimuthal distribution of the pressure fluctuations in the near wake of the spacer-grids for both configurations. The return to a homogeneous azimuthal distribution of the pressure fluctuations is faster without mixing vanes than with mixing vanes. The intensities of the pressure fluctuations are (in average) higher for WMV than for NMV by 13% and 136% at distances 0.5D h and 20D h from the grids respectively. The correlation functions and energy spectra highlight significant differences in the spatial and temporal structure of the pressure fluctuations for the configurations with and without mixing vanes. For both configurations, pressure fluctuations spectra reveal a frequency peak in the near wake of the grids. These peaks are around f D h / V flow ≈ 0.8 (69 Hz). Maxima are found at f ∗ = 0.78 (66.8 Hz) and f ∗ = 0.87 (74.6 Hz) respectively for WMV and NMV configurations. This frequency peaks persist farther from the grid for NMW than for WMV. The integral scale of pressure fluctuations highlight to different behaviour. The pressure integral length scale stays small compared to D h (ranging from 0.13D h to 0.2D h) for the NMV configuration whilst it increases with the distance from the grid for the WMV case. This is coherent with the persistence of frequency peaks for the NMV configurations. The pressure integral length scale is found to be 2–9 times larger for WMV than for NMV with the increase of the distance from the grid between 0.5D h and 20D h. A significant anti-correlated domain appears for NMW and not for WMV. In addition, LDV results show oscillations of the velocity profiles and a damping with the distance from the grids. Between 2D h and 15D h , the dimensionless ratio of "pressure fluctuations to velocity fluctuations" 〈 P rms ' 〉 θ / 1 2 ρ 〈 v rms ' 2 〉 P s q is found to increase by about 70% and 100% respectively for the configurations NMV and WMV. This increase further reveals the complexity of the turbulent flow in the wake of spacer-grids. [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]