Rod Bundle Research Articles

Thermal-hydraulic study in a wire-wrapped 19-rod bundle based on an isotropic four-equation model

The thermal–hydraulic performance of rod bundles with wire spacers is significant for studying Lead–Bismuth Eutectic (LBE) cooled fast reactor. The problem is that using the traditional constant turbulent Prandtl number of 0.85 ∼ 0.9, which is generally derived from the Reynolds analogy hypothesis suitable for conventional fluids, will greatly affect the numerical heat transfer accuracy of LBE having low Prandtl number properties. Thus, a great quantity of development work of four-equation models, which may improve the calculation accuracy of turbulent heat transfer in LBE flows, has been made by researchers, while no commercial code is available. The present work addresses an effort to study the thermal–hydraulic phenomenon in an LBE-cooled 19-rod bundle with wire spacers geometry on a self-compiled CFD solver (iso4eqnFoam), which is implemented by a standard finite volume method and C++ program and developed on OpenFOAM. In particular, the simulation aims to perform fully-developed turbulent stresses and turbulent heat fluxes by solving an isotropic four-equation model, which allows simple zero-fixed-value boundary conditions for turbulent variables and considers the difference between viscous and thermal boundary layer of LBE. The simulation results on a bare 19-rod bundle are first compared with heat transfer experimental data and other simulations to validate the effectiveness of the solver. Then, an LBE-cooled 19-rod bundle with wire spacers is presented with thermal–hydraulic simulations. The result shows that iso4eqnFoam can simulate the turbulent heat transfer of LBE under complex geometry. Finally, the distributions of temperature, temperature fluctuation, and turbulent Prandtl number in the channel of the wire-wrapped rod bundle are reported based on the isotropic four-equation model.

Validation of thermal–hydraulic system code improved by advanced drift-flux correlation using bundle data at moderate pressure conditions

The accurate prediction of the interfacial drag force is essential to predict the void fraction in a reactor core during the transient and accident phases of nuclear reactors. The key to accurate prediction is to provide reliable drift-flux parameters in calculating the interfacial drag force. The authors previously developed a drift-flux correlation to improve the interfacial drag force prediction accuracy under low-flow and low-pressure conditions. In the present study, the authors demonstrated the validity of the interpolation scheme at a moderate pressure range in the drift-flux correlation and quantified its uncertainty using the experimental data measured in a 5 × 5 rod bundle for the pressures of 0.25, 0.41, 2, 5, and 7 MPa at the multi-purpose Hitachi utility steam test leading facility (HUSTLE). The validated drift-flux correlation was implemented into the RELAP5/MOD3 code and numerical analyses were also performed to reproduce the HUSTLE experimental data. The numerical calculations using the modified RELAP5/MOD3 code showed improved prediction accuracy compared to the calculations by the original RELAP5/MOD3 with the Chexal-Lellouche correlation (e.g., EPRI correlation) in the tested pressure range. The uncertainty of the distribution parameter in the drift-flux correlation was estimated using the HUSTLE database. Next, a numerical analysis was done for the differential pressure in the reactor core in a small break LOCA integral test conducted with the ROSA/large scale test facility (LSTF) by the former Japan Atomic Energy Research Institute. This analysis showed that the predictions with the improved RELAP5/MOD3 agreed with the ROSA/LSTF data within the uncertainty at a 95 % confidence level. It was concluded that the improved RELAP5/MOD3 with the new drift-flux correlation could predict the water level of a reactor core during accident scenarios.

Experimental measurements of fluid flow in an 84-pin hexagonal rod bundle with spacer grid for a gas-cooled fast modular reactor

A 50 MWe helium-cooled fast modular reactor (FMR) is under development, and as part of the Department of Energy Integrated Research Project (IRP), Texas A&M University is conducting the thermal-hydraulic characterization of its baseline core configuration. We experimentally investigated the axial and cross flow fields characteristics in the hexagonal fuel rod bundle composed of 84 rods, a central rod, and spacer grids at a Reynolds number of 12,000. A fully transparent experimental facility resembling of one unit of the fuel assembly was constructed. Time-resolved particle image velocimetry (TR-PIV) measurements were performed to characterize hydraulic behavior downstream the spacer grid. Velocity measurements were conducted in the axial and radial direction of the rod bundle. From the PIV velocity vector fields, the full-field flow statistics were computed for the mean velocity, vorticity, and Reynolds stresses. An analysis of the energy decay downstream of the spacer grid was conducted with calculations of secondary-flow intensities, turbulent kinetic energy, power spectrum, and spatial-temporal velocity cross correlations. The vorticity field was obtained and pairs of counter-rotating vortexes were identified using a 3D reconstruction of several measurement planes. The data from this experimental campaign will be used to validate numerical models based on Computational Fluid Dynamics and other codes.

Computational Fluid Dynamics Modeling of a Pressurized Water Reactor Fuel Assembly to Estimate Loss Coefficients in Support of Subchannel Thermalhydraulics Modeling of Pressurized Water Reactor Small Modular Reactors With Advanced Fuels

Abstract Pressure loss coefficients are generally required by subchannel and system thermalhydraulics codes. These coefficients are not readily available for small modular reactors (SMRs) featuring nonconventional designs and novel coolants. In this study, the pressure loss coefficients were obtained using three-dimensional (3D) computational fluid dynamics (CFD) modeling for an advanced water-cooled reactor. A representative light water fuel assembly used in the Organization for Economic Co-operation and Development (OECD)/National Research Council Canada (NRC) pressurized water reactor subchannel and bundle tests (PSBT) benchmark was selected for CFD modeling and simulation under various working conditions. The fuel assembly includes three types of pressurized water reactor (PWR) spacer grids: simple grid (SG), nonmixing vane grid (NMVG), and mixing vane grid (MVG). Turbulent flow through subchannels of both nonheated and heated rod bundles was simulated to predict recoverable and nonrecoverable pressure distribution along the length of the bundle. It was observed that vortices were generated at the tips of spacer grids, affecting the cross-flow in subchannels significantly. The estimated pressure loss coefficients were found to be influenced by the flow conditions (Reynolds number or the upstream flow history) and spacer grid configuration. Pressure loss coefficient values ranged from 1.14 to 1.80, depending on the spacer grid type, design, and flow conditions. The CFD method used in this study was demonstrated to have the potential to generate input parameters required for the subchannel analysis and optimization of fuel assembly designs and serve as a surrogate for empirical correlations.

Proposal of a turbulent Prandtl number model for Reynolds-averaged Navier–Stokes approach on the modeling of turbulent heat transfer of low-Prandtl number liquid metal

Because of their high molecular heat conductivity, low-Prandtl number liquid metal is a promising candidate coolant for various designs of advanced nuclear systems such as liquid metal–cooled fast reactors and accelerator-driven sub-critical system (ADS). With the fast-growing computational capacity, more and more attention has been paid to applying computational fluid dynamics (CFD) methods in thermal design and safety assessment of such systems for a detailed analysis of three-dimensional thermal–hydraulic behaviors. However, numerical modeling of turbulent heat transfer for low-Prandtl number liquid metal remains a challenging task. Numerical approaches such as wall-resolved large eddy simulation (LES) or direct numerical simulation (DNS), which can provide detailed insight into the physics of the liquid metal flow and the associated heat transfer, were widely applied to investigate the turbulent heat transfer phenomenon. However, these approaches suffer from the enormous computational consumption and are hence limited only to simple geometrical configurations with low to moderate Reynolds numbers. The Reynolds-averaged Navier–Stokes (RANS) approach associated with a turbulent Prandtl number Prt accounting for the turbulent heat flux based on Reynolds analogy is still, at least in the current state in most of the circumstances, the only feasible approach for practical engineering applications. However, the conventional choice of Prt in the order of 0.9∼unity in many commercial computational fluid dynamics codes is not valid for the low-Prandtl number liquid metal. In this study, LES/DNS simulation results of a simple forced turbulent channel flow up to a friction Reynolds number Reτ of 2000 at Pr of 0.01 and 0.025 were used as references, to which the Reynolds-averaged Navier–Stokes approach with varying Prt was compared. It was found that the appropriate Prt for the RANS approach decreases with bulk Peclet number Peb and approaches a constant value of 1.5 when Peb becomes larger than 2000. Based on this calibrated relation with Peb, a new model for Prt used in the RANS approach was proposed. Validation of the proposed model was carried out with available LES/DNS results on the local temperature profile in the concentric annulus and bare rod bundle, as well as with experimental correlations on the Nusselt number in a circular tube and bare rod bundle.

The blockage model of Upgraded Cheng and Todreas correlation for pressure drop in hexagonal wire-wrapped rod bundles

A model to predict the pressure drops in a hexagonal wire-wrapped rod bundle with blockage is developed. The model is based on the Upgraded Cheng and Todreas Detailed (UCTD) correlation. Two series of experiments from Texas A&M University (TAMU) and Karlsruhe Institute of Technology (KIT) are used for calibration and validation. The UCTD correlation for drag and swirl forces are modified to account the presence of the blockage. The characteristic velocity of the Reynolds number is revised. The pressure drop is well predicted in the turbulent flow regime with a mean error less than 3%. The friction factor of transition flow regime is an interpolation between laminar and turbulent. Due to the lack of data in laminar flow regime, the mean error in the transition regime is higher than 5%. Overall, the mean errors of the two validation cases are 2.5% and −0.6%, and the root mean square (RMS) errors are 3.2% and 7.2%. To improve the accuracy and extend the ability to predict complex blockage conditions such as blocked thickness or porosity, a variety and an amount of experimental data are needed.

Prediction of Critical Heat Flux for Subcooled Flow Boiling in Annulus and Transient Surface Temperature Change at CHF

The ability to predict critical heat flux (CHF) is of considerable interest for high-heat equipment, including nuclear reactors. CHF prediction from a mechanistic model for subcooled flow boiling in rod bundles still remains unsolved. In this paper, we try to predict the CHF in an annulus, which is the most basic flow geometry simplified from a fuel bundle, using a liquid sublayer dryout model. The prediction is validated with both water and R113 data, showing an accuracy within ±30%. After the CHF in an annulus is calculated successfully, a near-wall vapor–liquid structure is proposed on the basis of the liquid sublayer dryout model. Modeling of heat transfer modes over the heating surface at CHF is performed, and predictions of the changes in liquid sublayer thickness and heater surface temperature at the CHF occurrence point are carried out by solving the heat conduction equation in cylindrical coordinates with a convective boundary condition, which changes with the change in flow pattern over the heating surface. Transient changes in the liquid sublayer thickness and surface temperature at the CHF occurrence point are reported.

Pressure Drop in Seven-Pin Wire-Wrapped Rod Bundle for the Sodium Cartridge Loop in Versatile Test Reactor

This work studies the hydrodynamics of the seven-pin wire-wrapped rod bundle in the sodium cartridge loop for the Versatile Test Reactor (VTR) through scaled water experiments and computational fluid dynamics (CFD) simulations. The scaling analysis is first performed to demonstrate the hydrodynamic similarity between water and sodium flows at the same Reynolds number . A separate-effects test facility is designed and constructed based on the scaling analysis. Detailed experimental data on the pressure drop covering a wide range of values (1165 to 27 689) are obtained, which are used to evaluate existing correlations for friction factor and to benchmark CFD simulations. The experimentally determined friction factors agree well with the Upgraded Cheng and Todreas Detailed Correlation and Pacio-Chen-Todreas Detailed Model within but are significantly underpredicted by Rehme’s correlation by 25%. Various CFD near-wall treatment methods are tested using ANSYS Fluent and evaluated by experimental data. It is found that when the recommended wall values are met, most of the near-wall treatment methods can give accurate friction factor predictions. The resolved near-wall method () with the Shear Stress Transport turbulence model and the scalable wall functions () with the realizable turbulence model can predict within . The standard wall functions () and nonequilibrium wall functions () with the realizable model can predict within ± 10%.

Uncertainty analysis and visualization for a coupled thermohydraulics-neutronics simulation by deterministic sampling

Uncertainty quantification and sensitivity analysis by deterministic sampling is performed on a thermohydraulics-neutronics coupling simulation of a 5×5 rod bundle in pressurized water reactor. Thermohydraulics calculation is conducted by a CFD simulation and neutronics calculation is conducted by method of characteristic (MOC). The coupling simulation of CFD and MOC, together with large number of calculations needed by random sampling in uncertainty quantification, leads to high computational cost. Therefore, deterministic sampling is introduced in this study and proves its efficiency. Moreover, the calculated results usually have some distribution pattern on the selected section, using mass average value as output is unable to reflect the distribution feature. Therefore, a method realized by Python program is introduced to perform uncertainty quantification and sensitivity analysis of the outputs on the outlet section in fine-mesh level. It is found that the uncertainty of outputs and the inputs’ impact on outputs vary in different areas at the outlet section.

Mitigation of jet cross-flow induced vibrations using an innovative biomimetic nozzle design inspired by shark gill geometry

Shark gill slits enable sharks to eject the water after the oxygen has been removed in ram ventilation. The reduced effect of jet flow from the gill slits gives sharks smooth maneuverability. A biomimetic “shark nozzle” is proposed to improve mixing between the jet flow and surrounding fluid. Jet flow systems are an essential component of many industrial applications. The key characteristic of jet flow is the mixing process that occurs between two fluid streams to allow heat and/or mass transfer between them. Many industrial and propulsion devices that use jet flows need rapid mixing for effective and environmentally friendly operation. Fuel-injection systems, chemical reactors, and heating and air conditioning systems are examples of devices where a mixing process takes place. Recently, jet flow plays an important role in design and operation of specialized nuclear pressurized water reactors (PWRs). The loss-of-coolant accident (LOCA) holes and slots machined in the core periphery baffle plates are designed to mitigate the effects of a severe LOCA event. However, in normal operation, the holes are a source of a jet flow that can induced vibrations in the fuel assemblies near the baffle before being mixed with the surrounding fluid. This may cause wear and fretting in fuel rods with their supports. The ultimate solution to prevent the fuel assembly vibrations from LOCA hole jetting in a reactor is enhancing mixing of a jet flow with ambient flow in order to rapidly reduce jet momentum. This work proposes a new shark-inspired nozzle design that exploits the observed high efficiency capacity of shark gill slits. Tests are conducted to evaluate the performance of the new design. The obtained results show that the shark-inspired biomimetic nozzle has a greater effect on the rod bundle vibration, and the critical velocity at which the unstable vibration occurs in the rod bundle is delayed by 20% using the biomimetic nozzle. In addition to delaying instability, a vibration amplitude reduction of 85% was obtained by using the proposed shark-inspired nozzle instead of the circular nozzle.

Recurrent neural network-based prediction of critical heat flux in rod bundles with non-uniform axial power shape

The Critical Heat Flux (CHF) prediction is one of the most important phenomena for the safety analyses of nuclear reactors. Current state-of-the-art CHF correlations for PWR fuel assemblies are based on classical regression approaches. In this paper a new method based on Recurrent Neural Networks (RNNs) for the calculation of the CHF in fuel assemblies is proposed. The sequential nature of the RNN allows it to learn the influence of the geometrical features of the fuel assembly and of the history effect by calculating the CHF not only based on the local flow conditions at the CHF location, but also upstream from it. The model was tested on two rod bundle datasets. The results show that the neural network model can reasonably well predict both the value of the CHF and its location without the need of any additional correction factors.

Numerical Investigation of Special Heat Transfer Phenomenon in Wire-Wrapped Fuel Rod of SFR.

Sodium-cooled reactors (SFR) have always been recognized as one of the most promising candidates for the fourth-generation nuclear systems as announced by the Generation-IV International Forum. In the design of SFR, helical wire-wrapped rod is applied to stabilize the structure of the rod bundle and enhance coolant mixing. Although there has been considerable research on SFR in computational fluid dynamics (CFD), the phenomenon of heat transfer has rarely been paid attention to. This article discovered that there exists reversed heat flux from coolant to wrapped wire, which is contrary to our usual understanding. This phenomenon has not been reported in previous CFD calculations. Hence, a solid heat conduction model is proposed to prove this phenomenon and analyze the heat transfer process. The simulation results show that the wrapping wire embedding depth, the shape of the calculation domain and the physical properties of all components have great influence on the magnitude of the reversed heat flux. The present findings will have strong influence on the temperature field and maximum value of the fuel rod as well as profound reference value for future flow calculation, especially in grid generation and treatment of the junction between the winding wire and fuel rod.

Modeling of Lateral Flow in CFD for Subchannel Analysis

This paper is aimed at the application of computational fluid dynamics (CFD) calculations for lateral flow modeling in rod bundles of Russian-type pressurized water reactors with hexagonal fuel rod lattice by subchannel analysis under a constant temperature. The subchannel code SUBCAL and CFD code ANSYS Fluent with the Reynolds stress turbulence model, which is capable of solving the anisotropic flow present in rod bundles, are used. Both methods are compared in terms of calculations in rod bundles. The literature review of available experiments of rod bundles suitable for CFD calculation validation follows. This paper describes the created CFD models on a triangular lattice, which are subsequently validated on selected experimental data in a wide range of Reynolds numbers and geometry (pitch-to–rod diameter ratio) together with mesh sensitivity analysis. The main part of this work is to develop a new equation for the lateral flow resistance coefficient for the subchannel code based on CFD calculations. Within these calculations, the turbulent mixing coefficient β for hydraulically smooth rod bundles, which is related to the geometry, and the momentum-energy transfer analogy correction factor ε are also evaluated and for which the equation is subsequently proposed.

Development of a nuclear fuel rod element positioning measurement equipment based on spacers retrieval network

The fast and precise positioning measurement of nuclear fuel rod element are an important part of the nuclear fuel rod bundle production line. This paper develops a new positioning measurement imaging device and proposes a positioning measurement algorithm based on spacer retrieval network. Firstly, the spacers retrieval network is trained to obtain the spacer’s types and location information, and statistically realize the type determination and initial positioning measurement of the nuclear fuel rod. Then, using the a priori information from the initial positioning, we intercept the regions of interest (ROI) of the spacers and perform contour extraction to achieve the precise positioning measurement of the nuclear fuel rod. The experimental results show that the equipment achieves 0.035° positioning measurement accuracy, and the positioning beat reaches 4.234 s/rod. The efficiency and accuracy of the proposed positioning equipment meet the application requirements of nuclear fuel rod bundle production line.

Numerical study on the effects of guide thimble on critical heat flux in a 5 × 5 fuel assembly

Rod bundle with guide thimble is a typical structure in the PWR fuel assembly. The thermohydraulics in the channel adjacent to the guide thimble are affected by the non-heated tube, introducing different characteristics in critical heat flux (CHF). In the current work, the CHF of 5 × 5 rod bundle with guide thimble at the center was predicted using the Eulerian two-fluid framework coupled with the critical heat flux model for heat transfer deterioration at the heated wall. The local thermal-hydraulic parameters as well as CHF values were compared with these without guide thimble. It was found that the CHF in the fuel assembly with thimble was slightly larger than that in the assembly without thimble. The lateral locations of boiling crisis were altered by the existence of the guide thimble. The inter-channel cross flow was also suppressed by the guide thimble due to its effect on the flow and thermal boundaries of the subchannels. The results obtained in this work can be used to support the safety analysis in the fuel assembly with guide thimble.

Study on the flow and heat transfer characteristics of liquid sodium in a hexagonal 7-rod bundle

The flow and heat transfer characteristics of liquid metal sodium are different from the ordinary fluids, which brings great challenges to the CFD simulation in the sodium-cooled fast reactor (SFR). Therefore, the accurate flow and heat transfer models are particularly significant in the high fidelity three dimensional analysis of liquid metal reactors. In this paper, the OpenFOAM solver implemented with k – ε – kθ – εθ turbulent heat transfer model was developed and used to study the flow and heat transfer characteristics of liquid metal sodium. Firstly, the proposed thermal hydraulic models were validated against the thermal hydraulic experiment of single-phase liquid sodium in a 7-rod bundle. The key thermal hydraulic parameters, such as the Nusselt number, pressure drop and frictional coefficient, were compared with the experimental data. Results show that the developed OpenFOAM solver with new implemented model performs well and the simulated results were in good agreement with the experimental data. Then the thermal–hydraulic characteristics of liquid sodium in the bend rod bundle were investigated. The effects of bending angle, Reynolds number on peak wall temperature and Nusselt number were investigated. CFD results suggest that the heat transfer in the bend rod bundle is highly sensitive to the bending angle but not to Reynolds number. Two opposite-direction cross flows were formed in the sections before and after the maximum bend position, which reveals the formation and disappearance of hot spot in the narrow gap.

Experimental study on the structure of the velocity boundary layer in a 3 × 3 rod bundle channel

The velocity boundary layer exists near the fuel rod surface in the fuel assembly due to the viscous effects. The boundary layer forms the main thermal resistance for heat transfer between the rod and the coolant since the momentum and energy transport is weak. In order to investigate the flow characteristics between the rod and the coolant in the rod bundle channel, this study directly measures velocity distribution in the boundary layer using the Particle Image Velocimetry (PIV) technique. The wall friction velocity is obtained by fitting the experimental data to the Spalding formula, choosing 0.436 as the Von Kármán constant, then the dimensionless velocity distribution of the boundary layer is obtained. The dimensionless velocity distribution under different Reynolds numbers is quantitatively studied. In addition, the effect of the gap size on the structure of the velocity boundary layer is investigated. The experimental results indicate that with the increase of the Reynolds number, the range of the inner layer of the boundary layer decreases gradually. In the inner layer of the boundary layer, with the increase of the Reynolds number, the proportion of the viscous bottom layer in comparison to the inner layer decreases, and the proportion of the logarithmic law layer in comparison to the inner layer increases, which enhances the flow and heat transfer performance in the channel. Reducing the gap size restricts the development of the velocity profile and increases the proportion of the inner layer of the boundary layer in comparison to the channel gap. A smaller gap size weakens the turbulence fluctuation, thus reducing the transverse mixing in the channel gap.

Modeling and Simulation Analysis on Mixing Characteristics of Two-Phase Flow Around Spacer Grid

The spacer grid has a significant impact on the thermal-hydraulic characteristics of the rod bundle channel, which directly determines the possibility of a boiling crisis. The multi-phase flow model coupling boiling model is proposed to acquire the secondary flow and void fraction for subcooled flow boiling in the fuel assembly. Compared with the experimental data, the RNG k–ε turbulence model is selected in the current research. In addition, the thermal-hydraulic parameters of subcooled boiling flow in the 5 × 5 rod bundle with a spacer grid are studied focusing on the influence of different arrangements of mixing vanes. It is found that the distribution of the vapor phase is greatly influenced by the intensity and flow direction of secondary flow. In addition, the thermal–hydraulic characteristics in the sub-channels and around the fuel rods are investigated, which demonstrates that uniform crossflow makes vapor harder to accumulate around the fuel rods. Furthermore, the mixing characteristics of different types of spacer grids are also discussed in detail.

Research on coherent structure in 4-rod bundle lattice based on CFD

Coherent structures exist only when the ratio of pitch to rod diameter (P/D) in the rod bundle lattice is less than 1.1. In order to produce coherent structures in large P/D lattices, roughness elements are arranged on the outer surface of the rod bundle. This method can enhance the heat transfer capability of the rod bundle lattice. The whole process of the generation and development of coherent structures are analyzed. The Reynolds stress model (RSM) is used for the numerical calculation of the flow and heat transfer of the working fluid in a 4-rod bundle lattice with a P/D of 1.25. The whole process of the generation and development of the coherent structures in the 4-rod bundle lattice is shown. The influence of different micro-rib size (e) on the coherent structures and the flow and heat transfer of the rod bundle lattice is analyzed. The interaction between the fluid channels in the 4-rod bundle lattice is also analyzed. The results show that with the arrangement of the micro-ribs in the large P/D rod bundle lattice, coherent structures are produced, which enhances the flow mixing between the channels of the 4-rod bundle lattice. The flow and heat transfer characteristics of the rod bundle lattice have been greatly changed, especially when P/(D + 2e) = 1.02, the heat transfer capacity of the rod bundle lattice increases by 44.1% while the flow resistance increases by 3.4 times. This research can provide theoretical support for the design and improvement of reactor fuel elements, heat exchangers and steam generators.

A multiscale model of a rod bundle using subchannel CFD

This paper deals with a coarse-mesh Computational Fluid Dynamics (CFD) approach, referred to as Subchannel CFD (SubChCFD), which combines features of the traditional 1-D subchannel analysis tools used in nuclear thermal hydraulic analyses and modern CFD. The SubChCFD, which has been previously developed for flow simulations, has been extended to perform thermal simulations and is then applied to a complex nuclear fuel bundle to explore and demonstrate the capacity of this new tool. SubChCFD allows resolved CFD sub-models to be nested into the regions that are of interest to locally improve the prediction, and can be coupled with the porous media approach to deal with any sub-scale fine structures that are difficult to be handled using a coarse mesh.In the present case studied, a SubChCFD baseline model is created, covering the entire heated length of the test geometry, to capture the axial developments of the flow and heat transfer in the rod bundle. Spacer grids that are used to keep the rods in place are modelled as embedded porous media to account for the associated blockage effect and pressure losses. In addition, a resolved CFD sub-model is created and coupled with the baseline model to improve the prediction for the region where high fidelity experimental measurements were performed. Through the present test, SubChCFD shows good predictability, flexibility and scalability in modelling large nuclear reactor components with complex internal structures. With the advanced coupling functionality, it is able to produce comparable simulation results to that of conventional CFD methods for regions of interest, with greatly reduced computational cost.