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Affected by sea wind and waves, the floating nuclear power plant (FNPP) would be in rolling motion during operation. The rolling motion will introduce fluctuated inertial force periodically, resulting in periodic thermal and hydraulic parameters variation, and may threaten the safety of the reactor. To investigate the influence of rolling motion on the heat transfer characteristics in the rod bundle channel under natural circulation condition, an experimental system with a 5×5 rod bundle channel was design and built on a rolling platform. An experimental investigation was conducted to study the influence of rolling motion on heat transfer in 5×5 rod bundle channels under rolling motion. The transient heat transfer characteristics of natural circulation in rod bundle were obtained. Experimental results demonstrate that the rolling motion induces periodic oscillations in the flow rate of the system. Furthermore, it is observed that with an increase in heating power, there is a reduction in the relative amplitude of these flow rate fluctuations. Consequently, this leads to periodic variations in Nu, accompanied by a corresponding decrease in relative fluctuation amplitude as well. The fluctuant flow disturbances disrupt the boundary layer and yield an enhancement of approximately 5% in the heat transfer coefficient. Regarding the local heat transfer characteristics, the incorporation of rolling motion introduces secondary flow which can significantly alter them. The presence of gravity and centrifugal force field fluctuations with a period half that of the rolling period leads to a local Nu curve with two peaks. With the center of the rod bundle channel as the axis, reverse phase fluctuations occur on both sides of the channel. The heat transfer performance is enhanced in subchannels near the rolling axis due to the rolling motion, but it diminishes away from that axis. The secondary flow induced by rolling motion amplifies velocity fluctuation amplitude in subchannels close to the rolling axis. Local Nu increases with an increase in power, whereas relative fluctuation amplitude decreases as power rises. The troughs of Nu are almost the same under different rolling periods, and the peak Nu fluctuates observed in subchannels become more pronounced as the intensity of rolling motion intensifies. In conclusion, this paper’s experimental results significantly contribute to advancing our understanding of heat transfer in rod bundle channels under oceanic conditions. The experimental results of this paper can provide experimental data for the study of wall temperature distribution of the rod bundle channel under rolling motion conditions.

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The gas-liquid two-phase flow widely exists in nuclear energy, chemical industry, petroleum engineering, and other industrial production fields. Two-phase flow will occur in pressurized water reactors during normal operation and accident conditions. The distribution of void fraction and interfacial area concentration in rod bundle channels has an important influence on the flow resistance, heat transfer, critical heat flux, and power distribution of the reactor. The complex geometric structure of rod bundle channels brings a great challenge to the local experimental measurement of two-phase flow. As a result, the existing database of the interfacial structure parameters in rod bundle channels is still insufficient, which greatly affects the accuracy of two-phase flow modeling. To clarify the transformation mechanism of the phase distribution in rod bundle subchannels and provide a database for future modeling, an experimental study of air-water two-phase flow in a 5×5 rod bundle channel was carried out in this paper. The four-sensor conductivity probe was used to measure the distribution of local two-phase flow parameters (void fraction, interfacial area concentration, bubble diameter, and bubble velocity) across the test section at 36.5 hydraulic diameters. The results show that with the increase of liquid phase velocity and the decrease of gas phase velocity, the “axial peak” gradually changes to the “wall peak” under the influence of radial force. The transition boundary of phase distribution is obtained and has a relatively accurate classification effect. The non-uniform void fraction distribution over the cross-section is observed, bubbles tend to gather in the central subchannels which are less affected by the wall and have a higher mass velocity. For the overall cross-section, the decrease of liquid phase velocity and the increase of gas phase velocity exacerbate the non-uniformity of bubble distribution. In the experimental conditions of this paper, the established correlation of void fraction and interfacial area concentration exhibits strong predictive capability. The average minimum relative error is ±18.2% and ±12.2% respectively. The drift flux model for the rod bundle channels should take into account various phase distribution types to enhance the modeling of distribution parameters. Furthermore, it is imperative to validate the model using comprehensive void fraction distribution databases to enhance the accuracy of prediction outcomes. Because of the small size of the bubble flow, the interfacial area concentration correlation established based on the vertical circular pipe database can also be used for rod bundle channels, but the modified correlation considering the influence of the bundle geometry on the bubble is more accurate. This study can provide a reference for the closure of the two-fluid model in the reactor thermal-hydraulic analysis program.

Fast and accurate evaluation of flow and heat transfer phenomena in rod bundles is a problem of long-standing interest in nuclear engineering. Computational fluid dynamics (CFD) can provide accurate but relatively time-intensive estimates, such that simulations of very long transients with high spatial detail are infeasible. On the other hand, subchannel codes require relatively low computational time and can provide pin-level estimates, but have substantial empiricism in evaluating aspects such as crossflows and turbulent mixing coefficients. A multiscale method (SC+) for bridging this accuracy/speed gap, based on the Subchannel CFD (SubChCFD) method of Liu et al. [Nucl. Eng. Design, Vol. 355, paper 110318 (2019)], is demonstrated and developed here with a focus on a 5 × 5 square rod bundle geometry. The method has been newly implemented into the commercial code STAR-CCM+ and benchmarked against Liu et al.'s data. The 5 × 5 geometry was chosen in part due to the availability of direct numerical simulation (DNS) data at a relevant Reynolds number of a similar configuration for comparison. The SC+ results are variously compared against results from DNS, large eddy simulation, wall-resolved Reynolds-averaged Navier-Stokes, and coarse-mesh CFD methods. As an expansion to the original SubChCFD approach, a simple Hi2Lo approach is demonstrated using the DNS data as a correction to the original friction factor correlations employed. This is verified to improve the predictions. Additional test cases with geometric perturbations are pursued, illustrating the flexibility of SC+. The potential of this method for modeling the narrow gap vortex instability, which would represent an advancement over standard subchannel approaches, is also assessed. The method is expanded to include transverse flow losses, which was found to improve the results for modeling the gap instability. Initial extensions of SC+ for hexagonal rod bundles are also presented; some inaccuracies for the coarsest meshes prompted a detailed investigation of the mesh convergence behavior of the method. Geometric correction factors were devised that provided substantial improvement on these very coarse meshes, improving the prospects of SC+ for wider usage. Future work plans are to expand the methodology to wire-wrapped rod bundles and to implement the method into Pronghorn, with a unified pipeline via the Cardinal wrapper between the codes NekRS, Pronghorn, and BISON, to solve fuel performance problems of direct interest to industry. [ABSTRACT FROM AUTHOR]

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The nuclear reactor system is complex and the operating environment is harsh, resulting in the complex phenomena of multi-physics coupling. The multi-physics coupling codes developed in the early stage shows limitations on codes' scalability and generality. Therefore, it is of great significance to build a multi-physics coupling framework and conduct research on key technologies in coupling problems, which may accelerate the development process of autonomous multi-physics coupling platform in China. In this paper, the multi-dimensional and multi-physics coupling finite element analysis platform for nuclear reactor developed by XJTU-NuTHeL was introduced. The main work consisted of the development of thermal-hydraulic model, the research of fuel performance analysis technology and the establishment of multi-physics coupling framework. In terms of thermal-hydraulic calculation, XJTU-NuTHeL conducted a series of studies on pressurized water reactors and advanced reactors grounded in the advanced multi-physics coupling framework, and developed the nuclear reactor system safety analysis code, NUSAC. In addition, a subchannel analysis model tailored for liquid metal fast reactors was established, and the fully coupled subchannel transient analysis code, FLARE, was developed. NUSAC and FLARE were then verified against relevant codes and experimental data. In the realm of fuel performance analysis, considering the wide application of finite element method in solid mechanics and its versatile modeling capabilities, XJTU-NuTHeL developed a fuel performance analysis code, BEEs, based on finite element method. The code could not only conduct multi-physics coupling analysis for traditional rod fuels under steady and transient conditions, but also extends its applicability to accident tolerant fuels and other fuels with diverse geometric shapes. This paper focused on the study and analysis of coated particle dispersed fuel and plate type fuel. The multi-scale simulation results of coated particle dispersed fuels, as well as the thermomechanical and corrosion behavior of plate type fuels were shown. In the context of multi-physics coupling analysis, the efficiency and accuracy of different mesh grid mapping schemes were studied and a multi-physics coupling framework was established. An example of the framework was then presented, showcasing the integration of the fuel performance code BEEs, the Monte Carlo neutron physics code OpenMC, and reactor system safety analysis code NUSAC. The keys parameters of mechanics, thermal-hydraulic and neutronics were obtained and analyzed through the coupling different codes. The multi-dimensional and multi-physics coupling finite element analysis platform built in this paper can provide a strong support for the high-fidelity numerical simulation of nuclear reactor multi-scale and multi-physics coupling.

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

High-resolution two-phase flow data in the rod bundle are important in the development and validation of high-fidelity models for computational fluid dynamics and subchannel codes, in particular, those pertaining to light water reactor cooling systems. The Michigan Adiabatic Rod Bundle Flow Experiment (MARBLE) has been constructed as a modular assembly of an 8 × 8 lattice rod bundle to simulate scaled pressurized water reactor and boiling water reactor subchannel assemblies. To establish a high-spatial resolution database of the void fraction in the reactor fuel assembly geometries, tomographic measurements were performed with the High-Resolution Gamma-ray Tomography System, which was designed and built in house; the detector system has a spatial resolution of less than 1.0 mm using 240 LYSO (Lu1.8Y0.2SiO5) scintillators with a fan-beam array. In the present study, the local void fraction was measured with the MARBLE facility under various air-water flow conditions (jg = 0.04 to 0.85 m/s and jl = 0.12 to 0.77 m/s) covering from bubbly to cap-turbulent flows. The local void fraction was also successfully measured under nonuniform and asymmetric air bubble distribution conditions with an investigation of the effect of spacer grids and mixing vanes on void drift across subchannels. [ABSTRACT FROM AUTHOR]

Wireless communication over Multiple Input and Multiple Output (MIMO) channel achieve increased transmission rate by dividing the input stream into a multitude of parallel data streams which are transmitted in parallel. Precoding at the transmitter aims to decompose the channel into an uncorrelated multiple subchannels using the channel decomposition technique so that data streams will be sent in parallel independent subchannels. The proposed analysis is to analyze the performances of the MIMO precoder using various channel decomposition algorithms are compared and its computational complexity is analyzed. The channel decomposition schemes considered are Singular Value Decomposition, Geometric Mean Decomposition, LDLH, LU, Schur, QR and Jordan decomposition. The simulation and analytical results confirm that precoding for the MIMO channel decomposed by the QR scheme outperforms all the other precoding methods based on channel decomposition in the light of Bit Error Rate performance and involves a relatively lesser number of Floating Point Operations. [ABSTRACT FROM AUTHOR]

Pronghorn-SC is a subchannel code within the Multiphysics Object-Oriented Simulation Environment (MOOSE). Initially designed to simulate flows in water-cooled, square lattice, subchannel assemblies, Pronghorn-SC has been expanded to simulate liquid-metal-cooled flows in triangular lattices, hexagonal subchannel assemblies. For this purpose, the algorithm of Pronghorn-SC was adapted to solve the subchannel equations as they are applicable to a hexagonal wire-wrapped sodium-cooled fast reactor. Cheng–Todreas models for pressure drop and cross-flow models were adopted and a coolant heat conduction term was added. To solve these equations, an improved implicit algorithm was developed robust enough to deal with the numerical issues, associated with low flow and recirculation phenomena. To confirm the prediction capability of Pronghorn-SC, calculations and comparisons with available experimental data of 19- and 37-pin assemblies were performed, as well as other subchannel codes. Finally, a flow blockage modeling feature was added. This capability was validated for both water-cooled square sub-assemblies and sodium-cooled hexagonal sub-assemblies, using experimental data of partially and fully blocked cases. [ABSTRACT FROM AUTHOR]

Tight lattice rod bundles can increase equipment compactness and improve heat exchange efficiency, and they are widely adopted in heat exchange engineering. Understanding the flow characteristics of gas-liquid two-phase flow in tight lattice subchannels is essential for developing heat exchangers and fuel assemblies with these tight bundles. The drift-flux model (DFM) is a crucial two-phase flow model extensively used in subchannel analysis codes. Investigating the DFM in tight lattice subchannels benefits the advancement of these codes. For two-phase flow interfacial structures, significant two-group characteristics exist, with bubbles divided into small (group-one) and large (group-two) bubbles. The substantial differences in flow characteristics between these two groups provide a solid foundation for developing a two-group DFM. This study examined the two-group characteristics of two-phase flow in a tight lattice interior subchannel and developed a corresponding two-group DFM. The two-group correlations for the distribution parameters and drift velocities at the tight lattice interior subchannel level were proposed and verified using experimental data. The accuracy of the developed two-group DFM in predicting group-wise void fraction and gas velocity was also verified. The standard relative deviation between the model predictions and experimental data was 8.11 % and 13.1 % for group-one and group-two void fractions and 8.33 % and 11.7 % for group-one and group-two gas velocities, respectively. This indicates satisfactory accuracy for the newly developed two-group DFM in a tight lattice interior subchannel. • The distinct two-group flow structure in a tight lattice subchannel was discussed. • The two-group distribution parameters were modeled. • The two-group drift velocity models were experimentally validated. • The two-group drift-flux model for a tight lattice subchannel was established. • Results by this two group drift-flux model agreed with experimental data. [ABSTRACT FROM AUTHOR]

• A new code system DDC-3D for neutronics and thermal-hydraulics coupling analysis has been developed using the two-step method as neutronics calculation methods. • A detailed demonstration of the computational strategy in neutronics and thermal-hydraulics coupling analysis is provided. • The BEAVRS benchmark is applied to validate the DDC-3D code system. • The results of the benchmarks prove the accuracy and feasibility of the DDC-3D code system for multiphysics simulations. The direct whole-core calculations can provide accurate results and insights to the physics phenomena of the reactor. It can also capture the local effects of temperature and density fields on fuel depletion. However, the computational cost of the direct whole-core calculations is expensive. To compromise between computational cost and accuracy, the DDC-3D code system has been developed to perform neutronics and thermal-hydraulics coupling analysis. The DDC-3D code system couples the open-source codes DRAGON & DONJON based on two-step method and subchannel code COBRA-EN. The Picard iteration method is applied to ensure the stability of numerical calculation. Then the BEAVRS benchmark is used to validate the computational capabilities of DDC-3D code system. The critical boron concentrations, control rod worths and fission rate distributions are calculated in HZP condition. The results show a good agreement with measured data. The results demonstrate that the two-step method is applicable and valid for multiphysics simulations. For the result of HFP condition for cycle 1, the results also agree well with measured data, including the trend of the critical boron concentrations and power distributions throughout the cycle 1. Although the detailed thermal–hydraulic experimental values are not available, the thermal-hydraulics analysis of the hot fuel assemblies indicates that the calculation results are reasonable. In general, the results demonstrate the feasibility and accuracy of DDC-3D code system for neutronics and thermal-hydraulics coupling calculations and life cycle simulation of PWRs. [ABSTRACT FROM AUTHOR]

The SubChannel Module, referred to as SCM henceforward in this document, is a subchannel code within the Multiphysics Object-Oriented Simulation Environment (MOOSE) developed at the Idaho National Laboratory (INL). SCM, is able to model flow through liquid-metal cooled, wire-wrapped fuel pin sub-assemblies, ordered in a triangular lattice. This work extends the existing validation done for these flow types and geometries and introduces validation for the newly implemented capability to model flows within a deformed duct. The subchannel duct deformation modeling approach consisted of reproducing the deformed duct effect on the geometric parameters of the boundary subchannels, i.e., modifying surface area, wetted perimeter and pin to duct gap. Then, the model was validated using experimental data taken from Areva's fuel cooling test facility (FCTF). [Display omitted] • SCM can simulate, sodium-cooled, wire wrapped hexagonal fuel bundles in a deformed duct. • The deformation is modeled by adjusting the geometric parameters of the edge subchannels. • SCM is validated against experimental data obtained from FCTF. • SCM calculations agree well with the experimental measurements. • The effect in the temperature distribution is minimal but observable in the axial pressure-drop. [ABSTRACT FROM AUTHOR]

• Nu oscillation induced by wrapped wire are correlated with area change. • Effects of P/D and wire shapes are studied in three subchannels of the bundle. • D is kept as a constant to ensure fixed rated power of the fuel bundle. • Relation of circumferential T and Nu distributions with wire rotation are studied. • P/D of 1.29 and trapezoidal wire are identified to show best performance. This study develops a numerical model to investigate the effects of wire wrapping on the flow velocity and temperature distribution in a seven-rod fuel bundle cooling by lead–bismuth eutectic alloy. The enhanced cooling performance of the wire-wrapped rod is compared with the bare rod, and the oscillated local Nusselt numbers along the axial direction in the inter, corner and edge subchannels are analyzed. The Nusselt number oscillation is found strongly correlated with the area change in different subchannels induced by the wrapping wires. In order to reduce the temperature gradient along the axial direction and eliminate local hot spots, the impact of different pitch-to-diameter (P/D) ratios and wire geometries on the flow and heat transfer performances are examined while considering the fixed rated power of the rod bundles. For the central subchannel, a higher P/D ratio leads to significant fluctuation in transverse velocity, reduced temperature and increased Nusselt number. With P/D of 1.29, the average temperature of the central rod is reduced by 13.2 °C. The influence of P/D in the corner and edge subchannels are not significant due to wall effect. Among the wires with circular, rectangular and trapezoidal shapes, the trapezoidal wire shows the highest Nusselt number and the highest TSEF (Transverse Secondary Flow) value indicating the best heat transfer performance. [ABSTRACT FROM AUTHOR]

The nuclear industry is investigating the feasibility of transitioning from 18- to 24-month fuel cycles because of the positive impact it would have on the operational costs for the current fleet of light-water reactors. A challenge to making this change is the increased risk of fuel fragmentation, relocation, and dispersal (FFRD) due to the known potential for ceramic fuel to pulverize into fine particles at the higher discharge burnups. Previous work has been performed by the Nuclear Energy Advanced Modeling and Simulation program to assess FFRD risk in high-burnup cores using the BISON fuel performance code and a coarse mesh thermal hydraulics (T/H) solution for a loss-of-coolant accident (LOCA) using the TRACE system T/H code. Because of the importance of the T/H solution for FFRD assessment, this study seeks to investigate the impact of using higher-fidelity subchannel techniques for modeling of the LOCA transient. CTF was used to model a subregion of a high-burnup core that was depleted by the Virtual Environment for Reactor Applications (VERA) multiphysics core simulator. Both coarse-mesh and pin-resolved models were created in CTF, and a consistent coarse-mesh TRACE model was also developed to allow for benchmarking the code results. A large-break loss-of-coolant accident (LBLOCA) reflood transient was simulated using these three models, and results were compared. Results showed some consistent differences between the CTF and TRACE coarse models, including a higher peak cladding temperature (PCT) prediction in CTF and later quenching in CTF; however, the transient clad temperature behavior was similar, and these differences are likely due to post-critical heat flux heat transfer modeling differences and minimum film boiling temperature model differences. The pin-resolved results indicate that the PCT in the lumped model is often under-predicted by as much as 70 °C and that PCT occurs at a different location than the high-power pin in the assembly. The lumped model predicts a difference of 10 °C or less between the average and hot pins in the assembly, whereas the pin-resolved model predicts a range of over 100 °C. These results indicate that higher-fidelity T/H results may have an impact on predicted core behavior during LOCA, which may be important to consider when assessing FFRD risk. • Fuel performance depends on thermal hydraulic behavior during severe accidents. • The impact of using subchannel methods to predict loss-of-coolant behavior was investigated. • Using higher fidelity methods allows for a more accurate determination of clad surface temperature behavior. [ABSTRACT FROM AUTHOR]

The marine environment may change the force on the fluid and inevitably influence bubble behavior and the two-phase flow in the reactor core, which are vital to the safety margin of a nuclear reactor. To explore the effect of the marine motion on the flow and heat transfer features of subcooled flow boiling in the reactor core, the volume of fluid (VOF) method is employed to reveal the interaction between the interface structure and two-phase flow in the subchannel under rolling motion. The variations of several physical parameters are obtained, including the transverse flow, the vapor volume fraction, the vapor adhesion ratio, and the phase distribution of boiling two-phase flow with time. Sensitivity analyses of the amplitude and the period of the rolling motion were performed to demonstrate the mechanisms of the influence of the rolling motion. We found that the transverse flow in the subchannel was mainly affected by the Euler force under the rolling motion. In contrast to the two-phase flow in the static state, the vapor volume fraction and vapor adhesion ratio show different characteristics under rolling motion. Additionally, the onset of significant void (OSV) point changes periodically under rolling motion. [ABSTRACT FROM AUTHOR]

Subchannel codes have been widely used for thermal-hydraulics analyses in nuclear reactors. This paper details the development of a novel subchannel code within the Idaho National Laboratory's (INL) Multi-physics Object Oriented Simulation Environment (MOOSE). MOOSE is a parallel computational framework targeted at the solution of systems of coupled, nonlinear partial differential equations, that often arise in the simulation of nuclear processes. As such, it includes codes/modules able to solve the multiple linear and nonlinear physics that describe a nuclear reactor, under normal operation conditions or accidents. This includes thermal-hydraulics, fuel performance, and neutronics codes, between others. A MOOSE-based subchannel code is a new addition to the fleet of INL-developed codes, based on the MOOSE framework. In this work, we present the derivation of the subchannel equations for a single-phase fluid, we proceed with the description of the algorithm that is used to solve these equations and describe how this algorithm was implemented within MOOSE. We also present how this code can be coupled to the BISON fuel performance code. Next, we verify the friction model and the turbulent mixing model. We calibrate the turbulent modeling parameters for momentum mixing and enthalpy mixing, C T , β . We validate the code using experimental results and last demonstrate the coupling capabilities using a simple example. [ABSTRACT FROM AUTHOR]

Wire-wrapped hexagonal fuel bundles have been extensively investigated due to their enhanced heat transfer and flow characteristics. Experimental measurements are important to study the thermal-hydraulic behavior of such assemblies and to validate and improve the predictive capabilities of specialized correlations and computational tools. Presently, very limited experimental data is available on the local subchannel pressure drop. Experimental measurements of subchannel pressure drop were conducted in a 61-pin wire-wrapped rod bundle replica, for Reynolds numbers between 190 and 22,000. Specialized instrumented rods were utilized to measure the local pressure drop and estimate the subchannels' friction factor. Three interior subchannels, one edge subchannel, and one corner subchannel were selected to study the effects of location and flow regimes on the friction factor and hydraulic behavior. The transition boundaries from laminar to transitions regimes, and from transition to turbulent regimes were estimated for the subchannels analyzed. The results were found to be in agreement with the predictions of the upgraded Cheng and Todreas detailed correlation (UCTD). The results of the experimental campaign provided a better understanding of the hydraulic behavior of the subchannels of wire-wrapped bundles, in relation to its geometrical features. [ABSTRACT FROM AUTHOR]

A direct reactor auxiliary cooling system (DRACS) under natural circulation (NC) conditions with a dipped-type direct heat exchanger (D-DHX) in the upper plenum of a reactor vessel (RV) has been investigated for enhancing the safety of sodium-cooled fast reactors. Studies of the past have revealed that core-plenum interactions, which consists of penetration of the coolant from D-DHXs into the subassemblies and the narrow gap between them (IWF: inter-wrapper flow), and the heat transfer through a wrapper tube among subassemblies (radial heat transfer), occurred and increased core cooling performance during the DRACS operation. Therefore, a multidimensional thermal-hydraulics analysis model in the RV using a computational fluid dynamics (CFD) code (RV-CFD model) was developed to evaluate the core cooling performance. For the design study, the RV-CFD model must simulate reasonable calculation costs while maintaining accuracy. In this study, the subchannel analysis method using the CFD code for fuel subassemblies (subchannel CFD model) was applied to the RV-CFD model. In the subchannel CFD model, the porous media approach was used to consider local geometry in the fuel subassembly, and the effective heat conductivity coefficients in a diffusion term of the energy equation were set to fit the actual radial thermal diffusion between subchannels. Two numerical simulations were compared to the experimental data obtained from the sodium experimental apparatus PLANDTL-1. In the first case, the focus was only the radial heat transfer without the D-DHX operation. In another case with the D-DHX operation, the IWF noticeably occurred, and the focus was on the core-plenum thermal interaction. The calculated sodium temperature in the core correlated well with the experimental results. The RV-CFD with subchannel CFD model was validated for core-plenum interactions during the DRACS with the D-DHX operation under NC conditions.

Pronghorn-SC is a subchannel code within the Multiphysics Object-Oriented Simulation Environment (MOOSE). Initially designed to simulate flows in water-cooled, square lattice, subchannel assemblies, Pronghorn-SC has been expanded to simulate liquid-metal-cooled flows in triangular lattices, hexagonal subchannel assemblies. For this purpose, the algorithm of Pronghorn-SC was adapted to solve the subchannel equations as they are applicable to a hexagonal wire-wrapped sodium-cooled fast reactor. Cheng–Todreas models for pressure drop and cross-flow models were adopted and a coolant heat conduction term was added. To solve these equations, an improved implicit algorithm was developed robust enough to deal with the numerical issues, associated with low flow and recirculation phenomena. To confirm the prediction capability of Pronghorn-SC, calculations and comparisons with available experimental data of 19- and 37-pin assemblies were performed, as well as other subchannel codes. Finally, a flow blockage modeling feature was added. This capability was validated for both water-cooled square sub-assemblies and sodium-cooled hexagonal sub-assemblies, using experimental data of partially and fully blocked cases.

• A newly subchannel analysis method is developed to model the bowed rod bundle geometry. • A newly developed CHF prediction method for rod bowing bundle is developed, which is independent of the empirical penalty factor and had a relatively strong versatility. • The effect mechanism of rod bowing on CHF is the variation of local parameters and the decrease of hydraulic-diameter. The penalty of fuel rod bowing on the departure from the nucleate boiling ratio limits must be considered in the thermal–hydraulic design. Different from the traditional penalty factor method, in this study, a relatively strong versatile method is proposed to predict the critical heat flux in rod bowing geometries by coupled subchannel analysis and mechanistic model. A subdivisional subchannel method is used to obtain more detailed local flow parameter and a distributed resistance model for rod bowing is developed to quantify the effect of rod bowing on local flow field. The bubble crowding model is coupled to the subchannel code to predict the CHF of the rod bowing bundles. The results showed that the present method can predict the CHF in rod bowing bundle well, even in the bowed-to-contact geometries. The effect mechanism of rod bowing on CHF is the variation of local parameter and the decrease of hydraulic-diameter. [ABSTRACT FROM AUTHOR]

• A mechanistic model for predicting CHF in rod bundles based on the high-precision subchannel code was developed. • A subchannel subdivision method and an incorporation of mixing vane grids crossflow models were implemented in subchannel code. • An equivalent-tube concept was proposed to convert the turbulent intensity distribution on the tube surface to that on the rod bundle surface. • The present model does not require any cold wall effect correction and can accurately predict the radial position of CHF. A mechanistic model for predicting critical heat flux (CHF) in rod bundles is developed based on the high-precision subchannel code ATHAS. To account for the non-uniform distribution of quality in subchannels caused by the presence of mixing vane grids (MVGs) and guide tubes (GTs), further subdivision of conventional subchannels and the incorporation of MVG crossflow models are implemented in ATHAS to provide a more detailed local flow field for the CHF model. The CHF mechanism model is developed based on the concepts of bubble crowding and assumption of local phenomenon. An equivalent-tube concept is proposed to convert the turbulent intensity distribution on the tube surface to that on the rod bundle surface. The present model is implemented in ATHAS and evaluated against CHF experimental data of rod bundles under the conditions of interest for pressurized water reactors (PWRs). The predicted results show good agreement with the experimental CHF data, with an average error and root mean square of 0.99 % and 4.69 %, respectively, for a total of 601 data points. Furthermore, owing to the use of the subchannel subdivision method, the present model does not require any cold wall effect correction and can accurately predict the radial position of CHF. [ABSTRACT FROM AUTHOR]

A statistic-volume averaged two-fluid model, convenient for subchannel and system codes in nuclear industry, is established from RANS equations of Neptune_CFD code and leads to the appearance of subgrid terms. A database of simulations is made on experiments with flow representative of nuclear reactor cores with cylindrical (KIT) and subchannel (PSBT) configurations. The subgrid terms are then compared to the convection terms on the database of 82 test cases. Concerning momentum and energy equations, subgrid terms reach respectively 7.5% and 16% of convection terms. Overall, the comparison shows that dispersion terms are always greater than turbulence terms, but both remain in the same order of magnitude, and are not negligible compared to convection. Moreover, the geometry has an impact on the dispersion terms. This observation suggests that caution should be exercised when using correlation established on a different geometry. • A macroscopic formulation is established from the RANS equations of Neptune_CFD. • A validated CFD database of subcooled boiling flows is created. • An a priori evaluation of subgrid terms is done based on CFD numerical simulations. • The subgrid terms are not negligible compared to the convection ones. • The channel geometry has an impact on dispersion terms values. [ABSTRACT FROM AUTHOR]

• Direct coupling of whole-core pin-by-pin neutronics and thermal-hydraulics codes. • The deterministic pin-by-pin neutronics code performs the master role. • The subchannel pin-by-pin thermal-hydraulics code performs the slave role. • Unified domain decomposition is designed for the parallelization of the two codes. • Block-wise ghost region is designed as a buffer between neighboring processors. In reactor design and safety analysis, the interaction between neutronics and thermal-hydraulics is of significant importance. As an alternative and improved two-step method, the pin-by-pin scheme requires pin-wise thermal-hydraulics feedback to improve the resolution of 3D whole-core analysis. In this work, we implement an embedded coupling of the subchannel code START with the whole-core pin-by-pin calculation system NECP-Bamboo2.0 following the master–slave approach. The 2D lattice code Bamboo-Lattice2.0 in NECP-Bamboo2.0 provides the pin-wise homogenized few-group constants for the coupling system, and the 3D whole-core pin-by-pin code Bamboo-Core2.0 is coupled as the master code with a modified MPI-based START. Bamboo-Core2.0 retains its neutronics module and multi-physics coupling strategy. In contrast, START is only embedded as a slave module into the master. Both of them share the same MPI-based parallelism strategy with a block-based domain decomposition approach. Therefore, the coupling code developed in this paper has a high-level global coupling efficiency on a multi-process platform. The data exchange between neutronics and thermal-hydraulics adopts a direct block-to-block model, thus requiring no additional data interface. The coupling code is verified using the VERA#6 3D single-assembly benchmark problem and the mini-core problem based on the VERA#4 benchmark. The numerical results demonstrate that the coupling code possesses good parallel efficiency and computational precision. Compared with the single-channel model, the subchannel model can simulate the mass/momentum/energy exchange between channels accurately, and thus a more continuous coolant temperature distribution can be obtained. Meanwhile, the subchannel model is also able to reduce the maximum pin-wise coolant temperature, fuel temperature, and power peak, while the eigenvalue can be increased by about 10 pcm for the steady-state problems used for verification in this paper. [ABSTRACT FROM AUTHOR]

• New method was developed to obtain subchannel average two-phase flow parameters. • New flow parameter mapping data was obtained for 8 × 8 rod bundle. • New two-group drift-flux model for 8 × 8 rod bundle was developed. • New two-group drift-flux model is useful for predicting interfacial area concentration. • Newly obtained data is useful for subchannel analysis code validation. Interfacial transfer terms in gas-liquid two-phase flows are formulated as the product of interfacial area concentration (IAC) and flux. The interfacial area transport equation (IATE) is essential for obtaining IAC in transient and developing two-phase flows. Two-group IATE was developed to account for the difference in the interfacial drag force between two groups of bubbles, where spherical and distorted bubbles are categorized as group one, while cap, slug, and churn turbulent bubbles are categorized as group two. The rigorous two-group approach requires two-group gas momentum equations, resulting in one additional momentum equation in the two-fluid model. The mixture gas momentum equation is considered to avoid the additional momentum equation. The two-group drift-flux model is necessary to calculate the two-group gas velocity from the mixture velocity. The present study develops an approximation methodology to acquire subchannel average and rod bundle average two-group two-phase flow parameters based on the validated power law assumption and a limited number of local data. The two-group model is developed for the distribution parameter and drift velocity for dispersed two-phase flows in rod bundles. The total performance of the newly developed two-group drift-flux model is evaluated by two-phase flow data in rod bundles. The results show that the developed two-group drift-flux model could well predict two-group gas velocities in rod bundles. This model is useful for subchannel thermal hydraulic analysis codes as a complement to the conventional one-group drift-flux model. [ABSTRACT FROM AUTHOR]

This paper investigates the admission control problem on the satellite multi-beam networks with non-orthogonal multiple access (NOMA). The goal is to maximize the number of supported users on the premise of ensuring the quality of service (QoS) by optimizing the subchannel and power allocation. We provide the system model and then formulate the admission control problem as a mixed integer non-convex optimization problem. The non-convexity and existence of integer variable make the optimal solution difficult to get. Therefore, we propose a joint subchannel matching and power allocation algorithm to obtain the suboptimal solution so as to reduce the computation complexity. The proposed algorithm can be used for both NOMA and orthogonal frequency division multiplexing access (OFDMA). Specifically, the subchannel matching problem is solved by a two-stage matching process where users are accessed to subchannel dynamically. The power allocation problem is modeled as a super-modular game where the existence and uniqueness of Nash equilibrium (NE) are analyzed. Moreover, an iterative power allocation algorithm is proposed based on the NE searching method. Finally, simulation results are provided for demonstrating the effectiveness and feasibility of the proposed algorithm.

The cooling of a nuclear reactor depends on a suitable fluid flow pattern among its fuel elements aiming the removal of heat produced in the fuel. In case of light water reactors, an excess of heat drives the fluid to change its phase from liquid to vapor, significantly reducing its capacity to remove heat and leading the reactor to a Loss of Coolant Accident. Numerical simulations using a CFD code is a suitable tool to address this kind of problem and explore the conditions that should be avoided during the reactor operation. The commercial CFD codes had proven to be reliable to simulate with a high accuracy and confidence the thermal-hydraulics of a sort of equipment and systems, avoiding spending efforts and financial resources in the development of new codes that, essentially, perform the same tasks. Despite of it, the CFD codes must be validated, such as against experimental results. To comply with this objective, a benchmark fuel element was purposed and experimentally essayed to provide experimental results for CFD codes calibration. The results of this essay are provided to the four types of subchannels for a 5x5 PWR fuel element, with results provided as density and void fraction. This work presentes the preliminary results obtained with CFD numerical simulations using the ANSYS-CFX® code for the central subchannel with active rods for stead state operation. The results demonstrated that the ANSYS-CFX® is adequate to simulate with high accuracy the flow in this subchannel.

A sodium-cooled fast reactor has been designed to attain a high burn-up core in commercialized fast reactor cycle systems. The sodium-cooled fast reactor adopts a wire spacer between fuel pins. The wire spacer performs functions of securing the coolant channel and mixing between subchannels. In high burn-up fuel subassemblies, the fuel pin deformation due to swelling and thermal bowing may decrease the local flow velocity in the subassembly and influence the heat removal capability. Therefore, understanding the flow field in a wire-wrapped pin bundle is important. This study performed particle image velocimetry (PIV) measurements using a wire-wrapped three-pin bundle water model to grasp the flow field in the subchannel under conditions, including the laminar to turbulent regions. The PIV results confirmed that the normalized flow velocity near the wrapping wire in the low Re number condition was relatively decreased compared to that in the high Re number condition. Meanwhile, in the region away from the wrapping wire, the maximum flow velocity was increased by decreasing the Re number. Accordingly, the PIV measurements using the three-pin bundle geometry without the wrapping wire were also conducted to understand the effect of the wrapping wires on the flow field in the subchannel. The results confirmed that the mixing due to the wrapping wire occurred, even in the laminar condition. These experimental results are useful not only for understanding the pin bundle thermal hydraulics, but also for the code validation.

BackgroundThe natural circulation chloride cooled fast reactor (N3CFR) has the characteristics of simple structure, high inherent safety and good economy. It is an advanced fourth-generation nuclear energy system with development potential. However, we are not able to calculate 3D thermal fluids in core when modeled using RELAP5-TMSR.PurposeThis study aims to improve the applicability and accuracy of the RELAP5-TMSR program in transient analysis and safety assessment of small natural circulation chloride cooled fast reactor (SN3CFR).MethodsFirstly, the coupling code was developed with an external explicit method, and verified by the horizontal tube model based on the system analysis code RELAP5-TMSR and the sub-channel code ThorSUBTH. Then, according to the natural circulation primary circuit main cooling system, a multiscale model of the SN3CFR was established to evaluate the applicability of the coupled code. Finally, the reactor steady-state operating parameters and the reactivity insertion incident conditions were calculated and analyzed.ResultsThe results show that the coupling code is in good agreement with the verification examples, and all key thermal parameters of SN3CFR meet design limits under reactive introduction accident conditions.ConclusionsThe development of a multiscale thermal hydraulics coupling code can perform system analysis fast and calculate the thermal fluid of the core more accurately, which is of great significance for the system design, safety analysis and optimization of molten salt cooled reactors.

The marine environment may change the force on the fluid and inevitably influence bubble behavior and the two-phase flow in the reactor core, which are vital to the safety margin of a nuclear reactor. To explore the effect of the marine motion on the flow and heat transfer features of subcooled flow boiling in the reactor core, the volume of fluid (VOF) method is employed to reveal the interaction between the interface structure and two-phase flow in the subchannel under rolling motion. The variations of several physical parameters are obtained, including the transverse flow, the vapor volume fraction, the vapor adhesion ratio, and the phase distribution of boiling two-phase flow with time. Sensitivity analyses of the amplitude and the period of the rolling motion were performed to demonstrate the mechanisms of the influence of the rolling motion. We found that the transverse flow in the subchannel was mainly affected by the Euler force under the rolling motion. In contrast to the two-phase flow in the static state, the vapor volume fraction and vapor adhesion ratio show different characteristics under rolling motion. Additionally, the onset of significant void (OSV) point changes periodically under rolling motion.

Subchannel codes have been widely used for thermal-hydraulics analyses in nuclear reactors. This paper details the development of a novel subchannel code within the Idaho National Laboratory’s (INL) Multi-physics Object Oriented Simulation Environment (MOOSE). MOOSE is a parallel computational framework targeted at the solution of systems of coupled, nonlinear partial differential equations, that often arise in the simulation of nuclear processes. As such, it includes codes/modules able to solve the multiple linear and nonlinear physics that describe a nuclear reactor, under normal operation conditions or accidents. This includes thermal-hydraulics, fuel performance, and neutronics codes, between others. A MOOSE-based subchannel code is a new addition to the fleet of INL-developed codes, based on the MOOSE framework. In this work, we present the derivation of the subchannel equations for a single-phase fluid, we proceed with the description of the algorithm that is used to solve these equations and describe how this algorithm was implemented within MOOSE. We also present how this code can be coupled to the BISON fuel performance code. Next, we verify the friction model and the turbulent mixing model. We calibrate the turbulent modeling parameters for momentum mixing and enthalpy mixing, CT,β. We validate the code using experimental results and last demonstrate the coupling capabilities using a simple example.

In order to join efforts to develop high-fidelity multi-physics tools for research reactor analysis, the KIT is conducting studies to modify the coupled multi-physics codes developed for power reactors. The coupled system uses the Monte Carlo Serpent 2 code for neutron analysis and the Subchanflow code for thermo-hydraulic analysis. Serpent treats temperature dependence using the target motion sampling method and Subchanflow was previously extended and validated with experimental data for plate-type reactor analysis. This work present for the first time the steady-state and transient neutron and thermo-hydraulic analysis of an MTR core defined in the IAEA 10 MW benchmark using Serpent2/Subchanflow. Important global and local parameters for nominal steady-state conditions were obtained, e.g., the lowest and highest core plate/channel power/temperature, the radial and axial core power profile at the plate level, and the core coolant temperature distribution at the subchannel level. The capabilities of Serpent2/Subchanflow to perform transient analysis with on-the-fly motion of the control plates were tested, namely with fast and slow reactivity insertion. Based on the unique results obtained for the first time at the subchannel and plate level, it can be stated that the coupled Serpent2/Subchanflow code is a very promising tool for research reactor safety-related investigations.

• Thermohydraulic calculation using Ansys CFX code. • Steady-state single-phase analysis. • Prediction of the Critical Heat Flux (CHF) and MDNBR assess. • Hottest subchannel analysis. • AP-Th 1000 core engineering design. This paper proposes a three-dimensional and single-phase model using computational fluid dynamics to study the thermal–hydraulic design limits of the AP-Th 1000 concept. This thermal–hydraulic study provides the engineering design limits of the hottest subchannel of the AP-Th 1000 concept during the first fuel cycle. The reactor successfully upheld a maximum temperature that consistently remained near the reference core during its operation. The AP-Th 1000 core has thermal limits comparable to those of other reactors in its technological class, such as AP1000. During the fuel cycle, the AP-Th 1000 concept maintains MDNBR values within the range of 2.26 to 1.97. The average values for Heat Flux HCF and Nuclear Enthalpy Rise HCF stand at 2.70 and 1.68, respectively. [ABSTRACT FROM AUTHOR]

Turbulent mixing effect is crucial for coolant subchannels in a nuclear reactor, as it directly influences flow distribution and cooling capacity. The enthalpy in the hottest channel is reduced in this way, thus improving the location of critical heat flux (CHF) to some extent. In order to predict CHF accurately, the inter-channel mixing model is incorporated into the subchannel analysis code to simulate thermal-hydraulic phenomena during a reactor operation. In this paper, a novel inter-channel mixing model has been developed through extensive investigation of experimental research. In addition, the power-law profile is considered for predicting the void distribution resulting from the void drift effect. Finally, the developed mixing model is embedded in a subchannel analysis code, and pressurized water reactor sub-channel and bundle tests (PSBT) data are selected as a benchmark for the model validation. The simulation results show that the current mixing model can accurately predict void fraction and thermal equilibrium quality at specific axial locations with absolute errors of 0.1 and 0.04, respectively. • A novel inter-channel mixing model that combines turbulent mixing and void drift effect has been proposed. • The influence of channel geometry has been taken into account in the current model. • The power-law method was used for characterizing void distribution resulting from void drift effect. • Current model accurately captures the axial development of parameters in different subchannel types. [ABSTRACT FROM AUTHOR]

• A whole-core thermal–hydraulic analysis code is parallelized and optimized to offer computation times compatible with pin-by-pin analyses. • Several parallelization and optimization strategies are proposed and implemented at the assembly-level and whole-core level, covering OpenMP, MPI and CUDA. • Parameters for evaluating the performance of the improved code are tested on an HPC cluster, and the speed improvements for each method are presented and analyzed. With increasingly developing high-performance computing (HPC) technology, pin-by-pin thermal–hydraulic simulations are now feasible. The in-house subchannel analysis code KMC-FBc has been extended to perform pin-by-pin calculations. To expedite the whole-core thermal–hydraulic analysis process, various parallelization and optimization strategies are proposed and implemented at the single-assembly and multi-assembly levels, including acceleration methods for solving large sparse matrix equation systems, memory and file input/output (I/O) optimization, hybrid MPI/OpenMP-based whole-core acceleration method and so on. Parameters for evaluating the parallel performance, such as the running time, speedup, and parallel efficiency, are tested on an HPC cluster. Results show that the maximum acceleration ratio for the whole-core problem with a mesh size in the tens of millions can reach approximately 120, with a parallel efficiency of around 32%, which demonstrates the rationality and effectiveness of the parallel algorithm and optimization methods and lays the foundation for future high-resolution and high-fidelity thermal–hydraulic analysis. [ABSTRACT FROM AUTHOR]

In the framework of the EU funded project NURESAFE, the subchannel code CTF and the neutronics code DYN3D were integrated and coupled on the NURESIM platform. The developments achieved during this 3-year project include assembly-level and pin-by-pin multiphysics thermal hydraulics/neutron kinetics coupling. In order to test this coupling, a PWR rod ejection transient was simulated on a MOX/UOX minicore. The transient is simulated using two different models of the minicore. In the first simulation, both codes model the core with an assembly-wise resolution. In the second simulation, a pin-by-pin fuel-centered model is used in CTF for the central assembly, and a pin power reconstruction method is applied in DYN3D. The analysis shows the influence of the different models on global parameters, such as the power and the average fuel temperature, but also on local parameters such as the maximum fuel temperature.

• In-depth interpretation of the thermal-hydraulics characteristics of flow blockage events in the fuel assembly. • Enhanced models for the wire-wrap spacer and convective heat transfer. • Good agreement between predictions and experimental data on velocity and temperature distributions downstream of the blockage. • Analysis of the impact of blockage location and size on coolant and cladding temperatures in fuel assembly of CEFR. The subchannel code ATHAS-LMR has been improved for analyzing thermal-hydraulics characteristics of flow blockage events in the fuel assembly of a sodium-cooled fast reactor (SFR). Based on the in-depth study of the phenomenon of flow blockage and combined with the characteristics of fast reactor assembly, enhanced models for the wire-wrap spacer and convective heat transfer have been added to account for changes in transverse flow in the presence of blockages. And appropriate transverse mixing and local resistance model are also implemented in ATHAS-LMR-FB. The improved code has been assessed against experimental data obtained at adiabatic and diabatic conditions in SFR assemblies with simulated flow blockages. Good agreement between predictions and experimental data on velocity and temperature distributions downstream of the blockage has been observed. Comparison of predictions of various analytical tools have also been performed. Predictions of the improved ATHAS-LMR code are shown to be compatible with those of other analytical tools. An in-depth analysis of effects of location and size of blockage has been carried out for the fuel assembly of the China Experimental Fast Reactor. Boiling has been predicted to occur for blockage ratios greater than 55.61% at the central location of the fuel assembly. Peak cladding and coolant temperatures are predicted at the axial location 70% of the rod length (just downstream of the maximum local power position of the non-uniform axial power profile). [ABSTRACT FROM AUTHOR]

Highlights • An open channel model is introduced and validated by experimental results. • The system stabilities under uniform, axial and radial non-uniform heating conditions are analyzed. • Effects of peak value, power peak location and power curve shape are analyzed under axial non-uniform heating condition. Abstract Two-phase flow instability is a very common phenomenon in two-phase systems, such as steam generator, BWR, boiler, condenser and so on. Normally, flow instability is not allowed to appear in the systems because it might result in potential issues such as thermal fatigue damage, mechanical vibrations, problems with system control as well as changing of heat transfer characteristics which may cause a boiling crisis. A lot of studies were presented on this subject based on uniform heating. However, non-uniform heating is the biggest characteristic in nuclear reactors which is different from most of the uniform heating systems. It is necessary to conduct research concerning non-uniform heating effect on two-phase flow instability for the reactor safety. This paper studies the flow instability in subchannels. The validation of subchannel model is performed using data from a GE 3 × 3 rod bundle test. The effects of peak value of power factor, peak location, and power curve shape along axial direction are examined. The stability comparisons based on stability boundaries are plotted with the phase change numbers versus subcooling numbers. The increase of peak value makes the system more stable when peak is at the downstream half of the test section and makes the system more unstable when peak is at the upstream side of the test section closer to the inlet. With power peak moves from inlet to outlet, the stability increases first and then levels off. With the increase of power shape factor F, which is the standard deviation of power curve data used to reflect power shape effect, stability increases when power peak is close to the exit and decreases when power peak is near the inlet of the test section. The radial non-uniform heating effect is not obvious in subchannel. According to the results of this stability study, recommendation is made on how to avoid conditions where two-phase flow instability easily occurs. [ABSTRACT FROM AUTHOR]

Highlights • The anisotropic turbulent mixing model for rod bundle with spacer grid is proposed. • Simulation result agrees well with experimental result with anisotropic turbulent mixing model. • Detailed thermal-hydraulic performance of rod bundle with spacer grid is analyzed. Abstract The interchannel mixing effect is very important for the thermal-hydraulic performance of fuel assembly subchannels in Pressurized Water Reactor (PWR), especially for that of the fuel assembly with spacer grid. In this study, an anisotropic turbulent mixing model with a turbulent mixing coefficient matrix is developed and embedded in an in-house code to analyze the thermal-hydraulic performance of the fuel assembly. The validation of the present turbulent mixing model has been performed by comparing the numerical results with the experimental results for the outlet temperature distribution of the fuel assembly. Then, the thermal-hydraulic performance of 5 × 5 rod bundle with spacer grid is analyzed by comparing the present turbulent mixing model and the conventional isotropic turbulent mixing model which uses a constant value to consider the average mixing effect. It is found that stronger diversion crossflow, more uniform outlet temperature distribution, higher Critical Heat Flux (CHF) and higher Departure from Nuclear Boiling Ratio (DNBR) of the fuel assembly are achieved when using anisotropic turbulent mixing model. In summary, the anisotropic turbulent mixing model depicts the interchannel mixing characteristics of spacer grid more reasonably. [ABSTRACT FROM AUTHOR]

Pronghorn-Subchannel, referred to as Pronghorn-SC throughout this document, is a subchannel code within the Multiphysics Object-Oriented Simulation Environment (MOOSE) developed at the Idaho National Laboratory (INL). Pronghorn-SC was initially designed to model flows in water-cooled, square-lattice, bare subassemblies. Its capability has been extended to model flows in liquid-metal-cooled, triangular-lattice, wire-wrapped subassemblies. To ensure the accuracy of Pronghorn-SC in predicting the behavior of liquid sodium-cooled reactors, the code was validated by comparing calculations with experimental data, obtained from the experimental breeder reactor (EBR-II), Shutdown Heat Removal Tests (SHRT) 17 and 45R. The steady-state calculation at the beginning of the transients was validated using temperature measurements taken at different axial elevations in the instrumented subassembly XX09. The validation exercise was performed in successive stages. First, a comparison between the measured temperature profiles and standalone Pronghorn-SC simulations using a uniform pin power profile was made. The pin power profile was then refined using a Serpent-2 model of the reactor core. Finally, the radial temperature profile was further corrected considering the inter-assembly heat transfer. A Pronghorn Finite-Volume (FV) thermal hydraulic simulation of XX09 and its six neighboring subassemblies; calculated the heat flux on the inner duct surface of the XX09 subassembly to inform the Pronghorn-SC model. Last, a transient validation calculated the peak temperature evolution during the SHRT tests. • Pronghorn-SC code can simulate sodium-cooled, wire wrapped, hexagonal fuel bundles. • Validation of Pronghorn-SC against EBR-II Shutdown Heat Removal Tests. • Pronghorn-SC calculations are enhanced by multi-physics/multi-scale coupling. [ABSTRACT FROM AUTHOR]

Studying thermal hydraulics in reactor cores is essential for ensuring the safe operation of nuclear power plants and advancing the development of reactors and fuel. In the CASL program initiated by the US Department of Energy, they selected the advanced nuclear reactor's numerical simulation method to establish a high-fidelity numerical reactor system, VERA. The thermal–hydraulic part of the VERA program's software system uses the subchannel program CTF for precise calculations and comprehensive analysis. Based on synthesizing and mapping more than 170 papers, this study provides a concise review of the present state in thermal hydraulics for light water reactors using the subchannel code CTF. The analysis method of the subchannel is discussed thoroughly, and the chronology of the development process of the subchannel program are provided. The physical model and simulation studies of CTF are also introduced, together with the investigations on the important phenomenon of thermal hydraulic treatment of the subchannel by CTF are summarized. For each of the above issues, the latest study results, existing difficulties and future trends will be presented. While subchannel analysis procedures have made great progress over the past few decades, existing subchannel codes' performance still faces challenges. We provide discussions and suggestions for further study of the CTF subchannel program to address these challenges. For CASL, CTF is the preferred program for analyzing thermal–hydraulic problems in subchannels of light water reactors. This review may potentially provide a useful reference for subchannel thermal–hydraulic analysis and improvement of CTF program. [ABSTRACT FROM AUTHOR]

A hexagonally arrayed 37-pin wire-wrapped rod bundle has been chosen to provide the experimental data of the pressure loss and flow rate in subchannels for validating subchannel analysis codes for the sodium-cooled fast reactor core thermal/hydraulic design. The iso-kinetic sampling method has been adopted to measure the flow rate at subchannels, and newly designed sampling probes which preserve the flow area of subchannels have been devised. Experimental tests have been performed at 20–115% of the nominal flow rate and 60°C (equivalent to Re ∼ 37,100) at the inlet of the test rig. The pressure loss data in three measured subchannels were almost identical regardless of the subchannel locations. The flow rate at each type of subchannel was identified and the flow split factors were evaluated from the measured data. The predicted correlations and the computational fluid dynamics results agreed reasonably with the experimental data.

Currently, safety analyses mostly rely on codes which solve both the neutronics and the thermal-hydraulics with assembly-wise nodes resolution as multiphysics heterogeneous transport solvers are still too time and memory expensive. The pin-by-pin homogenized codes can be seen as a bridge between the heterogeneous codes and the traditional nodal assembly-wise calculations. In this work, the pin-by-pin simplified transport solver Tortin has been coupled with a sub-channel code COBRA-TF. The verification of the 3D solver of Tortin is presented at first, showing very good agreement in terms of axial and radial power profile with the Monte Carlo code SERPENT for a small minicore and with the state-of-the-art nodal code SIMULATE5 for a quarter core without feedback. Then the results of Tortin+COBRA-TF are compared with SIMULATE5 for one assembly problem with feedback. The axial profiles of power and moderator temperature show good agreement, while the fuel temperature differ by up to 40 K. This is caused mainly by different gap and fuel conductance parameters used in COBRA-TF and in SIMULATE5.

The non-uniformity of circumferential heat transfer is small in conventional PWR core but becomes significant with tight-lattice core design for advanced reactors. Predicting the circumferentially non-uniform heat transfer behavior can be challenging given the considerable heterogeneity of the subchannel geometry and the drastic change of property with supercritical fluids. In this paper, a circumferentially non-uniform heat transfer model for subchannel analysis has been developed to predict the circumferential distributions of heat transfer coefficient, wall temperature and wall heat flux. In the model, the sources of the heat transfer non-uniformity are considered to be the circumferentially non-uniform flow area and the fluid property variation. To account for these two effects, new correlation with a non-uniform factor is developed. A series of tests using CFD method was performed for determining the empirical coefficients of the non-uniform factor. Furthermore, a two-dimensional fuel heat conduction model is also added to the subchannel analysis code. The new model was validated by comparing the prediction results with available experimental data of a 2 × 2 square rod bundle with supercritical water. It is demonstrated that the inclusion of circumferentially non-uniform heat transfer model leads to an improvement in the predictive capabilities for current subchannel analysis method and will improve the prediction accuracy of cladding temperatures in the design and safety analysis of reactor fuel elements. [ABSTRACT FROM AUTHOR]

Bubbly flow is a common natural phenomenon and a challenging engineering problem yet to be fully understood. More insights from either experiments or numerical simulations are desired to better model and predict the bubbly flow behavior. Direct numerical simulation (DNS) has been gaining renewed interests as an attractive approach towards the accurate modeling of two-phase turbulent flows. Though DNS is computationally expensive, it can provide highly reliable data for model development along with experiments. The ever-growing computing power is also allowing us to study flows of increasingly high Reynolds numbers. However, the conventional simulation and analysis methods are becoming inadequate when dealing with such ‘big data’ generated from large-scale DNS. This paper presents our recent effort in developing the advanced analysis framework for two-phase bubbly flow DNS. It will show how one can take advantage of the ‘big data’ and translate it into in-depth insights. Specifically, a novel bubble tracking method has been developed, which can collect detailed two-phase flow information at the individual bubble level. Due to the importance of subcooled boiling phenomenon in pressurized water reactors (PWR), the bubbly flow is simulated within a PWR subchannel geometry with the bubble tracking capability. It has been demonstrated that bubble tracking method significantly improves the data extraction efficiency for level-set based interface tracking simulations. Statistical analysis was introduced to post-process the recorded data to study the dependencies of bubble behavior with local flow dynamics. [ABSTRACT FROM AUTHOR]