Fuel Rod Bundle Research Articles

Modeling and Visualization of Coolant Flow in a Fuel Rod Bundle of a Small Modular Reactor

This article presents the results of an experimental study of the coolant flow in a fuel rod bundle of a nuclear reactor fuel assembly of a small modular reactor for a small ground-based nuclear power plant. The aim of the work is to experimentally determine the hydrodynamic characteristics of the coolant flow in a fuel rod bundle of a fuel assembly. For this purpose, experimental studies were conducted in an aerodynamic model that included simulators of fuel elements, burnable absorber rods, spacer grids, a central displacer, and stiffening corners. During the experiments, the water coolant flow was modeled using airflow based on the theory of hydrodynamic similarity. The studies were conducted using the pneumometric method and the contrast agent injection method. The flow structure was visualized by contour plots of axial and tangential velocity, as well as the distribution of the contrast agent. During the experiments, the features of the axial flow were identified, and the structure of the cross-flows of the coolant was determined. The database obtained during the experiments can be used to validate CFD programs, refine the methods of thermal-hydraulic calculation of nuclear reactor cores, and also to justify the design of fuel assemblies.

Core-winding of Sodium Cold Fast Reactor Based on CFD Method Numerical Study of Flow Characteristics

Abstract Liquid metal cooled fast reactor often uses spiral winding wire as the positioning component of fuel rod bundle. Besides fixing, spiral winding has the functions of strengthening subchannel mixing and heat transfer. These characteristics have a very important impact on the internal flow temperature characteristics of the fuel assembly and the safety of the core system, so it is necessary to conduct in-depth research on the flow field distribution of the fuel rod bundle channel with wire wound, the transverse and axial flow characteristics of the edge channel and its internal channel, the flow characteristics of the circulation and eddy current, the local temperature field, and the outlet temperature distribution characteristics. In the actual research, there are some problems such as high cost, difficulty and high risk to carry out thermal hydraulic characteristic experiment directly. This paper uses CFD fine simulation technology to study 7 rod bundles with wire winding, uses Fluent meshing to partition high-quality polyhedral unstructured mesh, uses ANSYS Fluent to conduct numerical simulation, studies the influence of the presence of rod bundles with wire winding on the distribution of flow field and local temperature field, and forms a mature and reliable simulation experiment scheme. The temperature change curve and lateral mixing phenomenon obtained by CFD simulation were investigated. The outlet temperature differences of side channel, inner channel and corner channel were investigated, and the convective heat transfer characteristics between channels were investigated due to the existence of wire winding.

Subcooled flow boiling characteristics and CHF in an eccentric annulus at low pressure conditions

Controlled flow boiling conditions enhance the heat transfer rate in various industrial applications such as heat exchangers, fuel rod bundles etc. In nuclear industry, the presence of mixing vanes, spacers, channel creep deformation etc., may lead to the formation of narrow and wide zones around the fuel rods. The asymmetry in the cross-sectional flow area around the fuel rod can be modelled as eccentricity which may cause an irregular coolant flow distribution affecting the heat transfer rate. In the present study, the eccentricity (e) is defined as the ratio of offset (s) to the difference in radii of inner and outer cylinders (Ro−Ri). The current study introduces a framework that involves calibrating the departure size of vapour bubble (D) in the EMF model framework which enables the accurate prediction of subcooled flow boiling characteristics and CHF power in an annulus (0.0≤e≤0.50) at low operating pressure conditions. The developed framework is validated with experimental data and a very good match is noticed. It is observed that eccentricity causes a decrease in coolant peak velocity magnitude by 12.8% in the narrow zone for e=0.50. In the subcooled region, the accumulation of vapour bubbles impedes further heat transfer to the liquid coolant, and thus the peak heated wall temperature (≈380.88K) is observed in the narrow zone for e=0.50. Eccentricity causes early coolant phase change which in turn results in early occurrence of CHF (q=522KWm−2) in the narrow zone of e=0.50. A quadratic relationship is also established to predict the CHF power in terms of eccentricity (e) within 5% variation.

Simulation of loss-of-coolant accident with thin-walled cladding tubes

A loss-of-coolant accident in a nuclear power plant was simulated in the CODEX-SLIM experiment. The electrically heated fuel rod bundle contained cladding material used at a VVER-440 power plant. The maximum temperature reached 900 °C and the experiment was terminated by water quenching. The condition of the bundle was investigated using non-destructive and destructive techniques. According to the experimental data, the fuel rods would maintain their integrity during a design basis loss-of-coolant accident.

Application of BPNN algorithm in thermal-hydraulic analysis of unwrapped LFR core

Due to the complex structure of the LFR fuel rod bundle, refined CFD simulation consumes significant computing resources and time, making it difficult to meet the immediate needs of engineering design and research. This paper introduces a method that integrates the sub-channel level thermal-hydraulic analysis code CorTAF-LBE with the BPNN algorithm to analyze the thermal-hydraulic phenomena within the LFR core. The method has been validated by refined simulation results of the THEADES test section, which indicate that the error is within an acceptable range. The feasibility of its application in LFR fuel assemblies was also evaluated. The flow characteristics and temperature distribution in unwrapped LFR rod bundles are investigated. Due to differences in fuel distribution and flow area, the temperature in the central channel is much higher than that in the side and trapezoidal channels. Three accident conditions, including inlet blockage, center blockage, and outlet blockage, were simulated. The corresponding peak temperatures of the fuel cladding were 670 K, 767 K, and 860 K, respectively, representing an increase of approximately 170 K compared to the nominal conditions. The analysis results effectively demonstrate the 3D temperature distribution of the fluid in the rod bundles, which is significant for reactor design and safety analysis of LFRs.

CorTAF-LBE: A full scale subchannel-level thermal-hydraulic characteristics analysis code for LBR core based on OpenFOAM

The lead-bismuth cooled fast reactor core adopts box-type design, and the fuel rod bundle is fixed by wrapping wires. The complex geometric structure of the fuel assemblies significantly affects the flow and heat transfer status of the liquid metal coolant. Building upon the open-source CFD platform, OpenFOAM, this work established conservation equations for liquid metal coolant, fuel rod thermal characteristic models, inter-wrapper coolant analysis models, and coupled heat transfer models, and developed the subchannel level thermal-hydraulic analysis code CorTAF-LBE for whole-core analysis of LBR core using the finite volume method. The wire-wrapped rod bundles internal flow and heat transfer experiments of lead-bismuth, carried out based on the THEADES and KYLIN-II circulation loops, were selected to perform verification and validation. The results match well with the experimental data and CFD fine-modeling prediction. The steady-state numerical simulation of three wire-wrapped rod bundle assemblies with inter-wrapper flow was carried out under normal operation condition. The temperature and pressure distributions of the coolant in each subchannel and inter-wrapper flow region, the temperature of cladding and other key thermal-hydraulic parameters were obtained. The role of the inter-wrapper coolant in heat removal from the fuel assemblies was revealed. The work in this paper has referential value for the optimization of lead-bismuth cooled fast reactor design and safety assessment during operation.

Study on safety gap between fuel channels in NHR200-II

This paper presents a novel method for estimating the safety gap between deformed fuel channels. The irradiation causes deformation of the fuel channel, which narrows the control rod channel, directly affecting the safety of the reactor. Investigating the critical safety gap between fuel channels can be utilized for safety assessments in design. Specifically, we first define the safety gap and three typical dropping cases. Next, we analyze the forces on the control rod, including flow resistance and channel resistance. Then, we establish a nonlinear dynamic model and simulate three dropping cases. Finally, we explore the influence of the friction coefficient and the fuel rod bundle. We find that the critical safety gap for the NHR200-II reactor is -5.0∼-3.1 mm (standard case) and -4.2∼-2.5 mm (arbitrary case), respectively. Our study provides a novel solution to accurately evaluate the effect of deformed boxed channels on the control rod’s dropping.

The reduced-order model of 5 × 5 fuel rod bundles

The fuel rod bundles are the core part of pressurized water reactors (PWRs), and its heat transfer characteristics directly impact the safety of PWRs. A computational fluid dynamics (CFD) model of 5 × 5 fuel rod bundles with a spacer grid is established, and the numerical simulation results are in excellent agreement with the experimental results. Then, the effects of four turbulence models, namely shear stress transport model, standard k–ε model, re-normalization group k–ε model, and realizable k–ε model on the thermal-hydraulic characteristics of the 5 × 5 fuel rod bundles are systematically investigated. Furthermore, two data-driven methods, namely proper orthogonal decomposition (POD) and dynamic mode decomposition (DMD), are used to analyze the flow fields of the 5 × 5 fuel rod bundles. The two methods can extract key modes or features to enhance the comprehension and description of the dynamic behaviors within the flow fields of 5 × 5 fuel rod bundles. Finally, two reduced-order models (ROMs), called the POD-radial basis function neural network surrogate model and DMD method, are constructed, which can enable rapid prediction of the flow fields for 5 × 5 fuel rod bundles with high accuracy. The CFD simulation results presented in this paper can provide valuable insights for studying the thermal-hydraulic characteristics of the 5 × 5 fuel rod bundles. The two ROMs proposed in this paper can significantly reduce the computational costs associated with studying the thermal-hydraulic characteristics of 5 × 5 fuel rod bundles.

Flow fields prediction for data-driven model of 5 × 5 fuel rod bundles based on POD-RBFNN surrogate model

The fuel rod bundles are a crucial component of nuclear reactors. Research the flow characteristics of the fuel rod bundles can be computationally expensive, despite some progress made in the past few decades. In this paper, the fast prediction of flow fields for 5 × 5 fuel rod bundles is implemented based on a data-driven algorithm. First, a refined model of the 5 × 5 fuel rod bundles with spacer grid is established, and meshes are generated using a hybrid meshes technology. Then, the flow characteristics of the 5 × 5 fuel rod bundles are extensively analyzed using Computational Fluid Dynamics (CFD) methods. A comparison between the CFD results and experimental data reveals a strong agreement, thereby validating our research findings. Subsequently, a reduced-order model (ROM) called the proper orthogonal decomposition (POD)-radial basis function neural network (RBFNN) surrogate model is proposed. the POD algorithm is utilized to extracting the dominant flow modes, and the RBFNN is employed to predict the POD mode coefficients of non-sample points. By linearly combining the predicted POD mode coefficients with the POD modes, rapid prediction of flow fields for non-sample points can be achieved. Compared to the full-scale CFD simulations, it finds that the proposed POD-RBFNN surrogate model not only significantly improved efficiency, but also maintained a high level of predictive accuracy. It is believed that the research results hold important value for CFD calculations within nuclear reactor cores.

Numerical investigation on critical heat flux for 5 × 5 fuel assemblies with a wide range of working conditions

To determine the safe boundary of fuel assembly operation and improve the heat transfer performance of fuel assembly, it is necessary to accurately calculate the thermohydraulic characteristics and the critical heat flux (CHF) in rod bundles. The Eulerian two-phase model coupled with extended wall boiling model was used to numerically investigate the 5 × 5 fuel rod bundle flows with four sets of mixing vane spacer grids. A wide range of working conditions were analyzed to validate the numerical method. The results indicate that the current numerical model can not only accurately calculate the CHF values but also predict the location of heat transfer crisis. Based on CFD (Computational Fluid Dynamics) results, the detailed two-phase flow distributions were provided to analyze the CHF performance under special structures, such as bowed rod and spacer backward shift. The quantitative results were obtained on the analyses of spacer position influence on CHF and the penalty effect of bowed rod. Furthermore, the prediction of present numerical model under low inlet subcooling conditions and cold rod CHF were discussed. The research of this paper provides support for the design and optimization of fuel assemblies.

Numerical investigation of flow stratification and heat transfer characteristics in a strongly heated horizontal rod bundles with helium-Xenon gas mixture

The helium-xenon(He-Xe) mixture has good heat transfer characteristics with a low Prandtl number(Pr ≈ 0.2 ∼ 0.3) and is suitable for the advanced compact gas-cooled reactor. The characteristics of horizontal flow stratification in the fuel rod bundles with He-Xe mixture under gravity and buoyancy were numerically simulated. Different wall heat flux q+ suitable for the reactor operating conditions were selected for calculation and analysis. Flow stratification is observed as strongly inhomogeneous in the radial direction. With the semi-local wall coordinates and Van-Driest transformation, laminarization is manifested by velocity and temperature profile and turbulent kinetic energy changes. In the horizontal flow, where the buoyancy is orthogonal to the direction of flow development, unique characteristics appear in the flow and heat transfer, and the growing longitudinal momentum and energy transport are considered the main reasons. The influence of flow acceleration and buoyancy under different bundle structures were compared. The buoyancy effect is evident in the loosely spaced rod bundles. In contrast, the thermally-induced flow acceleration effect in the tight-lattice scheme is noticeable.

Numerical simulation of flow and heat transfer in fuel assembly of floating nuclear power platform

ABSTRACT Floating nuclear power platform (FNPP), unlike onshore nuclear power plants, is subjected to periodic rolling loads at sea. These loads will have a more obvious impact on the flow characteristics and temperature distribution inside the fuel assembly (FA). This paper first establishes a refined model of the 7 × 7 core fuel assembly. Then, the flow and heat transfer characteristics of assembly with pure fuel bundles and assembly with guide tubes are analyzed using a segmented grid partitioning method, taking into consideration both accuracy and efficiency in calculation. Finally, the flow and heat transfer characteristics inside the assembly with guide tubes under rolling conditions are simulated and calculated, and compared with the previous simulation results. The calculation results show that the flow and heat transfer characteristics of the assembly with guide tubes are quite different from those of the pure fuel rod bundle assembly, and the rolling environment has obvious adverse effects on the flow field characteristics and temperature distribution in the fuel assembly of the offshore reactor, which must be considered in the design and construction of the floating nuclear power platform. The research results have significant reference value for the design and optimization of floating nuclear power platform assemblies under rolling conditions.

Numerical simulation to study mixing vane spacer effects on heat transfer performance of supercritical pressure fluid in an annular channel

The spacer represents an essential part in the nuclear fuel rod. Spacer grid with mixing vanes in fuel rod bundle of nuclear reactor core has a significant impact on heat transfer perfor-mance in downstream to grid spacer. Grid Spacers are located on the nuclear fuel rod as-sembly to hold suitable clearance among the rods in a bundle. The objective of this paper is to study the enhanced heat transfer performance of R134a at supercritical pressure 4.5 MPa near downstream to mixing vane spacer in a vertical channel of annular flow. A spacer of 0.38 blockage ratio with mixing vanes, situated at mid-span of an annular channel is used in the present work. Numerical simulations have been accomplished for spacer with mixing vane and spacer without mixing vane in an annular channel by using commercial CFD (Computa-tional fluid dynamics) code ANSYS Fluent. The present investigation represents the compara-tive study for spacer with mixing vane and spacer without mixing vane effects on heat transfer and flow field characteristics in a downstream direction for mass flow-rate 0.41469 kg/s and heat flux 160 kW/m2. The results indicate that spacer with mixing vane has notable influence on heat transfer performance and flow field characteristics downstream of mixing vane spacer as compared to spacer without mixing vane. Wall temperature fall and increase of coefficient of heat transfer are significantly greater adjacent to spacer downstream. Spacer influence in the improvement of the heat transfer is noted up to distance X/D = 40 downstream and then flow is found as fully developed.

Numerical investigation on characteristics of the onset of nucleate boiling in petal-shape fuel rod bundle assembly

The helical structure of the petal-shape fuel element results in a high heat transfer performance. Therefore, it is crucial to analyze its thermal–hydraulic characteristics. However, research on subcooled boiling in the petal-shape fuel rod bundle assembly is few. The characteristic of subcooled flow boiling in rod bundle assembly with a 2 × 2 petal-shape fuel rod was investigated numerically by Computational Fluid Dynamics. The results show that the wall superheat at the onset of nucleate boiling (ONB) decreases with the increasing mass flux and inlet subcooling degree, but has the opposite effect with pressure. The prediction accuracy of existing ONB correlations was investigated. The majority of correlations, however, deviate from the simulated values. Because they were not intended for the ONB of the petal-shape fuel rod, they did not account for the effects of enhanced heat transfer

Experimental investigation on distribution of boric acid in a vertical 1 × 2 rod array channel

During the reactor operation, the mixing mass flow of boron acid caused by diversion cross-flow, turbulent mixing and void drift changes the boron concentration distribution in the coolant and affects the amount of boron in the Chalk River Unidentified Deposit (CRUD). Accumulation of boron within the porous crud layer on the fuel rods in the cores of pressured water reactors (PWR) results in the crud-induced power shift (CIPS), threatening the operation safety of reactors. In order to obtain the mixing mass flow of boric acid between adjacent channels, a vertical test channel containing two subchannels simulating a PWR fuel rod bundle is constructed. Using this channel, flow distributions and mixing mass flow of boric acid are measured for single-phase and two-phase air–water flows. The measurements show that the boric acid distribution is affected by turbulent mixing for single-phase flow and the mixing mass flow of boric acid increases with the Reynolds number. For two-phase flow, the bubble-induced turbulence enhances the turbulence intensity of the liquid flow, the mixing mass flow of boric acid increases with the increase of void drift and reaches the maximum in the slug flow. The experimental data is used to support the development and validation of a new transverse source term model of boron mixing. The comparison result shows the significant improvement of the code prediction.

Characteristics of bubble behavior in petal-shaped fuel rod bundle channels under different flow conditions

The petal-shaped fuel element is a novel type of fuel element suitable for small-scale reactors. The motion of bubbles in the rod bundle channel has a significant impact on the surrounding flow and temperature fields, thereby influencing the two-phase flow and heat transfer characteristics. To investigate the dynamics of bubbles in the rod channel of the petal-shaped fuel rod, we simulated the rising motion of bubbles with different inlet flow rates using the volume of fluid method (VOF). We obtained the trajectory, rising speed, and shape changes of the bubbles with varying initial sizes, and analyzed the flow field around the bubbles. Our results indicate that the transverse displacement of bubbles in the radius range of 0.5 mm–1.25 mm increases gradually with their initial size, and the degree of deformation and maximum rising velocity of the bubbles also increase. However, bubbles with a radius of 1.5 mm experience less transverse displacement due to their larger size and contact with multiple fuel rods during the ascent process. Furthermore, the transverse displacement distance of the bubble increases and then decreases with the increase of fluid flow rate in the channel, while the aspect ratio of the bubble remains unaffected by the inlet flow rate. The study of bubble behavior sheds light on the mechanism of bubble motion in the channel of petal-shaped rod bundles, and provides a foundation for the study of two-phase flow and heat transfer in the channel.

Ultrasonic Velocity Profiler for the Measurement of a Bubbly Flow Velocity Vector in Small Channels

The multi-dimensional velocity distribution of coolant in bubbly flow within the fuel rod bundles of the reactor core in boiling water reactors (BWRs) is elucidated by experimental investigationin this study. Since a measurement technique is required for such an investigation, this paper proposes the development of an ultrasonic velocity profiler (UVP). The combination of special ultrasonic transducers and modified signal processing on the UVP is proposed to obtain a multi-dimensional velocity vector of the bubbles and liquid in bubbly flow. The ability of the proposed technique is demonstrated by performing an experiment in swirling bubbly flowand itsapplicability confirmed by comparing the resultswith another technique. The sound pressure distribution in the narrow channel of the rod bundle isthenmeasuredprior to the verification ofthe ultrasonic wave emitted through a small channel. The echo signal reflected from reflectors dispersed in the liquid,bubble, and tracer particles in the small channel of the rod bundle indicates that the proposed UVP can be applied inthis application with a low level of multi-reflection. Finally, the UVP system is demonstrated to measure the velocity vector of bubbly flow in the narrow flow channel on the rod bundle, andthe velocity vector of the bubble and liquid obtained simultaneously.

Fluidelastic instability of a fuel rod bundle subjected to combined axial flow and jet cross-flow

Nuclear fuel bundles are subjected to axial coolant flow in addition to jet cross-flow in a specific design of pressurized water reactors (PWRs). The combined axial flow and jet cross-flow was found to induce fretting wear of the fuel rods at the spacer grid position. Investigating a reduced fuel rod bundle under jet in transverse flow (JITF) condition is necessary to determine safe vibration free operation conditions for these reactors.This paper presents an experimental and theoretical framework to understand the fluidelastic behavior of a rod bundle subjected to axial flow and localized jet cross-flow. A 6x5 axisymmetric flexible single-span rod bundle is used to simulate a part of a fuel assembly under combined flow. Two working flow conditions, pure axial flow, and JITF are studied on a mock-up PWR assembly. The experiments show that the arrays are stable under pure axial flow as expected. Then, the responses of the rod bundles under combined flows (i.e. JITF) are measured at three test axial flow velocities, VAxial= 1.0 m/s, 1.5 m/s, and 2.0 m/s. The results show that the critical jet cross-flow velocity increases linearly with the axial flow velocity with a velocity ratio (VR = VJet/VAxial) of 1.6.The second part of the work is the development of a fluidelastic instability model to predict the stability boundary for rod bundles under combined axial flow and jet cross-flow. The effect of axial flow on the jet cross-flow is introduced by implementing an incidence angle (θ) of jet flow with respect to the rods. The fluid-added damping terms in the developed model are shown to be functions of θ. The fluidelastic forces are expressed as functions of the projected rod area derivative to account for rod vibration. The stability analysis shows the capability of the model to predict the trend of critical jet cross-flow velocity with the axial flow velocity. The model validation shows the predicted critical velocities to be within an absolute error range from 7% to 12.5% when compared to experiments. The theoretical analysis highlights the importance of the cross-coupling fluidelastic forces in predicting the instability of the mock-up PWR assembly.

PIV and CFD Study on Crossflow Characteristics in a 7-Pin Wire-Wrapped Bundle Channel

China initiative Accelerator Driven System (CiADS) combines a linac, spallation target and a Lead-cooled Fast Reactor (LFR) together, which is designed to transmute nuclear waste and accelerate the progress of China’s energy technology research towards the goal of carbon neutrality. A LFR uses helical wire-wrap spacers as positioning components to enhance crossflow mixing in the reactor core. To study the velocity distribution and crossflow characteristics in wire-wrapped rod bundle channels, a 2 : 1 magnified scale 7-pin bundle fuel assembly model was fabricated using polymathic methacrylate. Particle image velocimetry (PIV) and computational fluid dynamics (CFD) simulations were used to investigate the velocity distribution in the 7-pin bundle flow channels at Reynolds number of 1250~5000 in the x z plane and Reynolds number of 1500 and 2500 in the x y plane. The deviation between CFD simulation results and PIV experimental data was small, and the Reynolds Average Navier-Stokes model could accurately simulate the flow characteristics of the wire-wrapped fuel rod bundle channels. The maximum crossflow velocity caused by helical wires was about 40% of the axial bulk velocity. The normalized crossflow velocity at the subchannel interface varied approximately sinusoidally with the axial height. As the Reynolds number increased, the velocity distribution trend and the loss rate of axial velocity in flow channels remained essentially constant while the peak value of crossflow velocity increased. The contour images of velocities with different axial heights were obtained from the x y plane, and their velocity distribution had a certain periodicity. The axial velocity loss rate in each subchannel caused by wire-wrap spacer resistance was between 7.35% and 38.51%, and the axial velocity loss rates in inner subchannels were usually higher than those in edge subchannels.

Principal component analysis of vibration response of a rod bundle subjected to jet cross-flow

Flow injecting through loss of coolant accident (LOCA) holes in the baffle plates of some pressurized water reactors (PWRs) can potentially induce vibrations of reactor fuel assemblies. For reactor fuel rods in close proximity to LOCA holes, the jet flow transverse to the rod axis may lead to vibration levels significantly higher than those for rods located away from the LOCA holes. The jet cross-flow-induced vibration could, in turn, induce fretting wear of the fuel rod cladding at the fuel assembly supports, also known as spacer grids.Fuel rods have axisymmetric flexibility, testing the real condition of the fuel assembly is important to know the most dominant vibration direction relative to the jet flow. This information could be used to suppress the dominant vibration by increasing the stiffness of spacer grids in the dominant direction. In this study, the effect of jet cross-flow on a 6x6 square rod bundle axisymmetric vibration is experimentally investigated. This research aims to answer the following important question: Does there exist a generalized eigenvector model that could be used to predict rod bundle vibration in jet flows, in addition to characterizing the stability effect of the diameter ratio? The dynamical behavior of the rod bundle is studied for three jet-to-rod diameter ratios; 1.6, 2.3, and 3.0. Flow-induced vibration tests are performed to determine the stability threshold of the bundle as well as to extract the bundle vibration mode shapes for the three nozzles tested. The experimentally measured mode shapes identify the dominant vibration direction which provides valuable insight into the fluid–structure interaction dynamics. The experiments show that increasing the diameter ratio lowers the critical velocity at which fluidelastic instability occurs. The maximum vibration amplitude in the bundle, on the other hand, decreases as the diameter ratio increases.Analysis employing PCA and SVD confirmed the existence of a set of robust modes that exist for the range of system parameters tested. Hence, modes shape obtained with different nozzles are the same for the same acceleration ratio of the jet flow. The existence of these structurally stable (in parameter space) fluid–structure modes has significant implications for the stability analysis of reactor fuel rods subjected to jet cross-flow. The results show that reduced order models of the high-dimensional fluid–structure can be developed and that the models can realistically capture the dynamics of the jet cross flow-induced vibrations of the fuel rod bundle.