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AbstractReflood thermal hydraulics remains a difficult and complex subject, and understanding the physical phenomena that occur during a reflood transient is important to nuclear safety. The Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) Rod Bundle Heat Transfer (RBHT) project was designed to provide unique experimental data for code assessment and model development. Participants, which came from 21 international organizations, used analysis codes including APROS, ATHLET, CATHARE, CTF, MARS, RELAP5, TRACE, and SPACE to simulate the tests performed in the RBHT facility.The experimental campaign carried out within the OECD/NEA RBHT project produced data for a total of 16 reflood tests conducted in two test series. An “open” test series consisted of 11 experiments, and a “blind” test series consisted of 5 experiments. In the blind tests, only the initial and boundary conditions were provided to participants prior to simulation of those experiments. Reflood rates ranged from 0.5 to 15 cm/s, thus producing data applicable to dispersed flow film boiling and inverted annular flow film boiling. Inlet subcooling ranged from 2.8 to 80 K. Tests with variable reflood rates and oscillatory reflood rates were included in the test matrix. This paper describes the project and presents a summary of major experimental and analytical findings. [ABSTRACT FROM AUTHOR]

The critical heat flux (CHF) is a vital parameter influencing the safety and efficiency of reactor cores. In this study, the Eulerian two-fluid model coupled with the extended wall boiling model in STAR-CCM+ was employed to simulate the departure from nucleate boiling (DNB) phenomenon in a 5 × 5 pressurized water reactor (PWR) fuel rod bundle channel with spacer grids under non-uniform heating conditions. The transition in boiling curves was used as the criterion of DNB occurrence, while the temperature distribution of rod surfaces was utilized for CHF location predictions. The predicted CHF value and CHF location exhibited good agreement with the experimental data. The deviation between calculated and experimental CHF values was within 15% and the deviation between predicted and experimental CHF locations was within one grid-to-grid span length. The results of this study suggested good prospects for the application of two-phase CFD model in predicting CHF in fuel assemblies with spacer grids.

Spacer grids and mixing vanes exhibit a significant role in the thermal hydraulics of pressurized water reactors (PWRs), especially in the post loss-of-coolant accident regimes. A detailed analysis of the contrasting upstream and downstream turbulent flow features is of great importance to both system codes and computational fluid dynamics (CFD)–Reynolds-averaged Navier–Stokes (RANS) modeling. Further, with the advent of supercomputing resources and machine learning research, a data-driven approach to turbulence modeling is gaining popularity. However, owing to the complexities associated with large-scale, high-fidelity data collection capabilities, the application of machine learning–based turbulence models has been limited to simple geometries. In this work, using a highly scalable CFD code PHASTA, we have collected data from direct numerical simulations of a PWR subchannel with high spatial and temporal resolution. From the collected data we extract key turbulent flow features, including mean velocities and Reynolds stresses that highlight the effects of spacer grids and mixing vanes on downstream turbulence in a typical PWR subchannel. An invariant analysis of the anisotropic stress tensor is also presented, which further elucidates their effect on the nature of turbulence in the immediate upstream and downstream vicinity. The high-resolution data from the simulations are archived and intended for the development of data-driven RANS closure models that are capable of capturing the evolution of anisotropy in PWR subchannels. [ABSTRACT FROM AUTHOR]

High fidelity nuclear reactor fuel assembly simulation using CFD method is an effective way for the structure design and optimization. The validated models and user practice guidelines play critical roles in achieving reliable results in CFD simulations. In this paper, the international benchmark MATiS-H is studied carefully and the best user practice guideline is achieved for the rod bundles simulation. Then a 5 × 5 rod bundles model in the advanced pressurized water reactor (PWR) is established and the detailed three-dimensional thermal-hydraulic characteristics are investigated. The influence of spacer grids and mixing vanes on the flow and hear transfer in rod bundles is revealed. As the coolant flows through the spacer grids and mixing vanes in the rod bundles, the drastic lateral flow would be induced and the pressure drop increases significantly. In addition, the heat transfer is enhanced remarkably due to the strong mixing effects. The calculation results could provide meaningful guidelines for the design of advanced PWR fuel assembly.

Spacer grids are one of main components of a Pressurized Water Reactor (PWR) fuel assembly. They are able to improve heat transfer from rod bundles to the water flow by increasing turbulence and mixture of this flow. On the other hand the pressure drop increases because spacer grids. Experimental and Computational Fluid Dynamics (CFD) analysis have been used to understand how spacer grids affect the water flow. This analysis is important to improve spacer grids thermal-hydraulic performance. This paper aims to investigate numerically and experimentally the water flow through PWR spacer grids. The numerical and experimental procedures have been developed for a 5x5 rod bundle with spacer grids at the Nuclear Technology Development Center (CDTN) in Belo Horizonte, Brazil. At CDTN, measurements of the velocity components are acquired with a 2D LDV (Laser Doppler Velocimetry) system and the numerical results are obtained using ANSYS CFX code. The measurements are obtained at one height downstream from a spacer grid and compared to CFD simulations for a flow rate at Reynolds number of 5.4x104 . Results show good agreement between both methodologies. The great repeatability and low experimental uncertainty evaluated (< 1.24%) in this work can be used to validate other CFD codes.

The spacer grids are structures present in nuclear fuel assembly of Pressurized Water Reactors (PWR). They play an important structural role and also assist in heat removal through the assembly by promoting increased turbulence of the flow. Understanding the flow dynamics downstream the spacer grid is paramount for fuel element design and analysis. This paper presents water flow velocity profiles measurements downstream a spacer grid in a 5x5 rod bundle test rig with the objective of highlighting important fluid dynamic behavior near the grid and supplying data for CFD simulation validation. These velocity profiles were obtained at two different heights downstream the spacer grid using a LDV (Laser Doppler Velocimetry) through the top of test rig. The turbulence intensities and patterns of the swirl and cross flow were evaluated. The tests were conducted for Reynolds numbers ranging from 1.8x104 to 5.4x104. This experimental research was carried out in thermohydraulics laboratory of Nuclear Technology Development Center – CDTN. Results show great repeatability and low uncertainties (< 1.24 %). Details of the flow field show how the mixture and turbulence induced by the spacer grid quickly decays downstream the spacer grid. It is shown that the developed methodology can supply high resolution low uncertainty results that can be used for validation of CFD simulations.

Experiments aimed at studying the effect of fuel rod vibration in a turbulent coolant flow has on the effort of moving and longitudinal displacement of a fuel rod in the system of VVER-440 reactor fuel assembly (FA) spacer grids (SGs) for the case when the fuel rod is not fixed in the lower support are carried out. It has been found that the fuel rod longitudinal displacement force becomes essentially (by approximately 30%) higher in a coolant flow than it is in stagnant water. The fuel rod's longitudinal displacement force decreases with an increase in the water flow velocity and, hence, with an increase in the fuel rod bundle vibration intensity. At the water flow velocity equal to approximately 3.8 m/s, which is close to the its nominal value in the VVER-440 reactor FAs, the displacement force decreases by approximately 10% as compared with its value at the velocity equal to 0.8 m/s. The regularities relating to the change with time in the fuel rod's longitudinal displacement speed and acceleration testify that the force resisting its motion in the spacer grid system has a periodic pattern. During vibration (flexural oscillations) of a fuel rod bundle, the stiffness in the fuel rod–SG cell system changes its value, and the axes of SG cells shift relative to one another for the characteristic period of one vibration cycle. It is particularly the above-mentioned effects connected with fuel rod bundle vibration that lead to a growth in the fuel rod's longitudinal displacement force. The results from the accomplished calculated assessments show that the algebraic sum of the Archimedean force, gravity force, force due to pressure loss across the fuel bundle, and inertial forces with the water velocity values up to 4.81 m/s do not exceed 6% of the minimal fuel rod longitudinal displacement force. In turn, this means that the system of honeycomb-type spacer grids reliably secures the VVER-440 reactor fuel rods that are not fixed in the FA lower support grid. [ABSTRACT FROM AUTHOR]

The paper considers heat transfer and temperature fields in a free-packed fuel pin bundle cooled by a heavy liquid metal with different types of spacing. Data is analyzed for three fuel assemblies with a pitch of s/d = 1.33: a smooth fuel pin bundle, a bundle of fuel pins spaced by a bilifar-helix wire wrapper of the “wire-to-wire” type, and a bundle with transverse spacer grids. In the free-packed smooth fuel pin bundle, there are no temperature non-uniformities around the fuel pin periphery in contrast to major general temperature non-uniformities in the bundle with wire wrapping which provides for a major decrease in the heat transfer. In the bundle with transverse spacer grids, the heat transfer only increases in the region of the grids, and, between them, is approximately equal to the heat transfer in the smooth pin bundle. Correlations are presented as recommended for the calculation of Nusselt numbers and the temperature non-uniformities around the fuel pin periphery with the above spacing options.

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]

Efficient algorithms for pre-processing of eddy-current signals that are typical when inspecting the condition of metal in the heat-exchanger pipes of steam generators of NPPs with VVÉR-440 and VVÉR-1000 reactor facilities are described. Flaws in these pipes can be reliably detected and classified, with their geometrical parameters evaluated, if one can tune out from multiple interfering factors and disturbances that affect the signal and emerge due to a number of structural irregularities in the test object (geometrical noise (pilger noise), spacer and antivibration grids, bends), specific detector-design features, and noises in the measuring tract of eddy-current apparatus. The tuning-out algorithms have been tested with experimental signals and incorporated in the PIRATE (Program for Identification & Recognition of defects through signal Analysis in Testing by Eddy currents) software intended for analyzing multifrequency signals. [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]

• 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]