Search

This paper describes a numerical study of flow-induced vibration in a helical coil steam generator experiment conducted at Argonne National Laboratory in the 1980 s. In the experiment, a half-scale sector model of a steam generator helical coil tube bank was subjected to still and flowing air and water, and the vibrational characteristics were recorded. The research detailed in this document utilizes the multi-physics simulation toolkit SHARP developed at Argonne National Laboratory, in cooperation with Lawrence Livermore National Laboratory, to simulate the experiment. SHARP uses the spectral element code Nek5000 for fluid dynamics analysis and the finite element code DIABLO for structural analysis. The flow around the coil tubes is modeled in Nek5000 by using a large eddy simulation turbulence model. Transient pressure data on the tube surfaces is sampled and transferred to DIABLO for the structural simulation. The structural response is simulated in DIABLO via an implicit time-marching algorithm and a combination of continuum elements and structural shells. Tube vibration data (acceleration and frequency) are sampled and compared with the experimental data. Currently, only one-way coupling is used, which means that pressure loads from the fluid simulation are transferred to the structural simulation but the resulting structural displacements are not fed back to the fluid simulation. [ABSTRACT FROM AUTHOR]

• Rarely reported LES data are provided for heat transfer in HLM flow. • Pr t = 1.5 are suggested for RANS model in HLM simulation through the comparison with the LES data. • The local Pr t models are investigated and seems to be a promising approach if we can calibrate the coefficient properly. The Westinghouse Electric Company, together with an international team, is developing its next-generation high-capacity nuclear power plant based on lead-cooled fast reactor technology. In general, the Prandtl number of a heavy metal fluid is two orders of magnitude lower than that of water. Because of the low Prandtl number of lead, the thermal boundary layer in the temperature field is much thicker than the viscous boundary layer in the velocity field. Existing engineering turbulence models all use the Reynolds analogy for coupling temperature and velocity fields, but it is not valid for Heavy Lead Metal flow. Therefore, the selection of an appropriate turbulent Prandtl number (Pr t) is crucial for lead fluid simulation. The conventional choice of Pr t as a constant value of 0.9 is not valid for low-Prandtl-number fluid flow. This paper aims at using an advanced high-fidelity numerical approach, Large Eddy Simulation (LES), to provide detailed insight into the physics of the flow and the associated heat transfer in a rod bundle with liquid metal flows. The results for the global Nusselt number with different Pr t values (0.9 ∼ 3.0) indicate that a higher Pr t can reduce the predicted Nusselt number by up to 44.1 %. The detailed temperature distributions obtained with RANS and LES are compared to better understand the deviation introduced by the turbulence model. For turbulent flow, the correlations for Nusselt number proposed by Cheng and Tak (2006) are in good agreement with the simulation, especially for a Peclet number < 1000. The analysis shows that the RANS model with Pr t = 1.5 shows better agreement on the prediction of local temperature distribution and global Nusselt number. Although the modified local turbulent Prandtl number model shows some improvement over the constant Pr t model, the development of a more sophisticated local turbulent Prandtl number model is necessary for further applications. [ABSTRACT FROM AUTHOR]

Abstract Recent progress in data-driven turbulence modeling has shown its potential to enhance or replace traditional equation-based Reynolds-averaged Navier-Stokes (RANS) turbulence models. This work utilizes invariant neural network (NN) architectures to model Reynolds stresses and turbulent heat fluxes in forced convection flows (when the models can be decoupled). As the considered flow is statistically one dimensional, the invariant NN architecture for the Reynolds stress model reduces to the linear eddy viscosity model. To develop the data-driven models, direct numerical and RANS simulations in vertical planar channel geometry mimicking a part of the reactor downcomer are performed. Different conditions and fluids relevant to advanced reactors (sodium, lead, unitary-Prandtl number fluid, and molten salt) constitute the training database. The models enabled accurate predictions of velocity and temperature, and compared to the baseline k − τ turbulence model with the simple gradient diffusion hypothesis, do not require tuning of the turbulent Prandtl number. The data-driven framework is implemented in the open-source graphics processing unit–accelerated spectral element solver nekRS and has shown the potential for future developments and consideration of more complex mixed convection flows. [ABSTRACT FROM AUTHOR]

• Mixed convection phenomena for advanced reactor coolants is investigated. • Summary of direct numerical simulations, turbulence modeling and machine-learning based approaches is presented. • Detailed analysis of turbulence for various Pr number fluids is performed. The adoption of liquid metals and molten salts as coolant fluids in advanced reactor designs has challenged traditional turbulence and heat transfer models because of different convective heat transfer characteristics from water. The challenge is exacerbated by the yet-to-be-understood mixed convection regime, which is crucial to passive heat removal. The NEAMS IRP Challenge Problem 1 (CP1) project aims to facilitate scale-flexible heat transfer models that can serve for broader applications in reactor thermal–hydraulic analysis. CP1 research activities include direct numerical simulation (DNS) and the development of advanced turbulence modeling techniques. The DNS efforts study the low- and high-Prandtl mixed convection in the identified canonical flow scenario, from which the fundamental understanding on the effect of buoyancy on turbulence and heat transfer is investigated. The CP1 modeling endeavors aim at advancement of the present engineering models and the employment of the data driven (DD) methods for the turbulence modeling. In the engineering model concentration, the performance of the current CFD models is assessed in the non-unitary Prandtl flows to gain physical understanding of their shortcomings. A framework for model error prediction and error propagation estimation is also established. A DD RANS framework is developed for the frozen prediction of Reynolds stress and turbulent heat flux based on the polynomial tensor representation of the respective quantities. The DD framework showed satisfactory performance over the k - τ model in the forced and mixed convection cases. [ABSTRACT FROM AUTHOR]

A single-phase heat pipe flow model implemented in the heat pipe application Sockeye is described. This model solves one-dimensional, compressible flow equations for the vapor phase in the center of a heat pipe, which are coupled to the two-dimensional heat conduction equation for the wick, liquid, and cladding, as well as to an ordinary differential equation tracking the working fluid inventory in the evaporator section of the heat pipe. This model is demonstrated with several test problems, including comparisons to analytic solutions for the vapor flow fields, analytic curves for sonic and capillary limitations of heat pipe operation, and some experimental data. The numerical solution gives excellent agreement for verification problems and good agreement with experimental results. Also, demonstrations show that the model is very robust, allowing for full simulations of heat pipe transients, including frozen startup, sonic-limited (supersonic) flow, and heat pipe shutdown. • A novel heat pipe model was developed and implemented in the code Sockeye. • Liquid layers are modeled with heat conduction and the vapor with compressible flow. • A series of test problems demonstrates robustness and accuracy. [ABSTRACT FROM AUTHOR]

Computational Fluid Dynamics (CFD) has become increasingly important in the research and development of advanced nuclear reactors. In the current study, extensive CFD simulations were conducted for the coolant flow in Molten Salt Fast Reactor (MSFR) core models using the state-of-the-art spectral element flow solver Nek5000 and multiscale coarse-mesh thermal-hydraulic software Pronghorn. The underlying motivation is to seek an in-depth understanding of how the internal velocity distribution can be influenced by the MSFR core cavity shape, the Reynolds number, turbulence modeling options and the inlet boundary conditions. The CFD techniques involved in this investigation range from coarse-mesh CFD, RANS modeling, to the high-fidelity LES calculations. Specifically, a series of RANS simulations were performed for the 2-D axisymmetric core model and 3-D wedge domains to study the flow distribution inside the MSFR core. It is observed that a proper representation of the MSFR inlet channel duct is important for the prediction of internal flow distribution. It is also showcased here how researchers can leverage the Nek5000 CFD results to calibrate more efficient coarse-mesh CFD tools, like Pronghorn, for the actual MSFR design needs. Moreover, this paper highlights a 3-D LES model for an entire MSFR core using the spectral element method and demonstrates the feasibility of this modeling approach. The readiness and potential limitations of the RANS approach are examined with respect to the high-fidelity LES simulations. The present investigation lays a solid foundation as we are leveraging the high-fidelity CFD capabilities to inform MSFR design efforts. • Applications of various CFD techniques in modeling MSFR core flows. • RANS calculations of 2-D axisymmetric and 3-D wedge domains of MSFR. • LES simulations of a 3-D MSFR full model at modest Reynolds number. • Calibration of coarse-mesh CFD tool Pronghorn using Nek5000 datasets. [ABSTRACT FROM AUTHOR]

• Applying tet-to-hex meshing method to SEM CFD code Nek5000 for complex nuclear-related geometries. • Boundary layers were created to obtain good near wall resolution. • Several example cases, including fuel assemblies, helical coil steam generator, random packed pebble bed, are simulated with Large Eddy Simulation, showing the capability of tet-to-hex meshing method of handling complicated geometry. • In most of the presented cases, experimental data was presented along with numerical data. In general, a good agreement was obtained, validating the accuracy of the tet-to-hex meshing method. • Compare to an ill-designed block mesh, tet-to-hex mesh is more robust and numerical efficient. The spectral element code Nek5000 is an open-source, higher-order computational fluid dynamics code developed at Argonne National Laboratory. It is designed to solve incompressible Navier-Stokes equations, but it also has a low-Mach-number approximation feature available. Large eddy simulation is approached by explicit filtering of the velocity field (and other fields) to mimic the effect of dissipation due to the unresolved scale. The computational domain is decomposed into second-order hexahedral elements that conform to the boundaries. However, generating a high-quality pure-hexahedral mesh can be challenging for some problems. For simple geometries, traditional blocking methods can be used to decompose the domain into smaller blocks to generate a so-called structural mesh. A structural mesh can maintain good orthogonality but can have a highly skewed mesh to conform to the geometry, as well as unnecessary refinement in the far field. Moreover, for geometries with relative complexity, blocking the geometry becomes impossible. To address these issues, we adopted a tet-to-hex strategy to generate a pure hexahedral mesh for Nek5000. First, we generate a pure tetrahedral mesh for the geometry; then we divide one tetrahedral element into four hexahedral elements. A pure tetrahedral mesh could be easily generated for complex geometries by using many current meshing codes. In this paper, we use the commercial codes ANSYS meshing and ANSYS ICEM to generate the pure tetrahedral mesh and then convert it to a pure hexahedral mesh. Boundary layers are extruded in ANSYSICEM to maintain near-wall resolution. [ABSTRACT FROM AUTHOR]

Pebble-bed reactors (PBRs) rely on a continual feed of fuel pebbles being cycled through the core. As a result, they require a "run-in" period in order to reach an equilibrium state. The run-in period for a PBR is a complex, time-dependent problem that requires the injection of new fuel, different types of fuel, and power increases. This complexity in the run-in makes it important to capture the physical processes in order to generate an accurate representation. The present work details the creation of a high-fidelity Monte Carlo methodology for analyzing the run-in and subsequent approach to equilibrium for PBRs. The methodology entails a Python module wrapped around Serpent so as to perform neutronics calculations, move pebbles, refuel the core, and discharge pebbles, thereby modeling the explicit behavior of the PBR run-in. Three run-in simulations (a constant temperature profile, a linear temperature profile, and a constant temperature profile using control rods) were examined in order to identify the key physical phenomena present in the run-in process. Utilizing kugelpy , we found the inclusion of a temperature profile to be important for accurately capturing a discharge burnup (around 141 MWd/kg), a consistent k-eff (around 1.005), and an average pebble power (around 2.5 kW/pebble) that all fall within acceptable limits. • Simulation of the run-in to equilibrium phase of a pebble-bed reactor. • Examination of isothermal and a simplified temperature profile during the run-in. • Examination of how control rods affect the run-in process. [ABSTRACT FROM AUTHOR]

• Novel approach to model turbulent thermal stresses in highly symmetrical geometries. • Application to triangular array rod bundles. • Full-height LES performed for the present project and presented for comparison. • Results are superior to available RANS and vastly cheaper than full scale LES. • Further speedup obtained with a novel Steady State solver in the SEM code Nek5000. An alternative methodology is proposed here to overcome the excessive cost of large eddy simulations (LES) of full-length heated rod bundle calculations, while improving the inaccurate results typically obtained with Reynolds-averaged Navier–Stokes equations (RANS). While the cost of the full-length LES is generally too high, LES of a small section of a single rod is usually affordable. The idea presented here consists of using the information granted by the small LES calculation to determine the appropriate turbulence viscosity or turbulent thermal diffusivity that can be used to solve only for the temperature field in a pseudo-RANS approach. The study has been performed with single-rod simulations with a P / D of 1.12 and 1.24, considering rod lengths that are representative of reactor applications, for the cases of uniform heat flux and a more realistic cosine-like axial heat distribution. The spectral element code Nek5000 has been used for all LES, RANS, and pseudo-RANS simulations. The recently proposed Nek5000 steady-state solver has been used for solving the temperature field in the pseudo-RANS approach and has proved significantly faster than transient schemes. Prediction of thermal quantities is compared with classical linear and nonlinear RANS models. LES for the full-length rods has also been performed and is used as a reference. Results of the proposed method show significant improvements with respect to those obtained with RANS. [ABSTRACT FROM AUTHOR]

Nuclear system thermal hydraulic analysis has historically relied on computationally inexpensive 1D codes. However, such tools are unable to capture multiscale multidimensional effects in large nuclear reactor enclosures. On the other hand, simulations with higher fidelity can be too expensive for such purposes. One of the ways to reduce computational cost is to perform simulations on a coarse grid, which, unfortunately, introduces large discretization errors. In this paper, two high-to-low data-driven approaches are investigated: (1) a coarse grid turbulence model to predict eddy viscosity and (2) correction of errors in coarse grid velocity fields. The approaches aim to reduce grid- and turbulence model-induced errors in coarse grid Reynolds-averaged Navier–Stokes (RANS) simulations. Two sources of high-fidelity data, RANS and large eddy simulations (LES), are explored. To extract the eddy viscosity from the LES data, an inverse optimization problem is solved. However, the LES eddy viscosity is shown to be comparable to the RANS eddy viscosity in terms of error reduction. Therefore, the directly available RANS eddy viscosity was used to develop a coarse grid data-driven turbulence model. Additionally, error correction in velocity is used to reduce the remaining uncertainties and bring the results closer to reality. The performance of the frameworks is demonstrated for a scaled upper plenum of a gas-cooled reactor facility. • High-to-low data-driven methods to inform coarse grid simulations are explored. • Coarse grid data-driven turbulence model is developed using LES and RANS data. • Inverse optimization problem is solved to extract eddy viscosity from LES. • Data-driven error correction in solution is applied. • Significant improvements of the low-fidelity coarse grid results are observed. [ABSTRACT FROM AUTHOR]

This study summarizes the experimental effort to characterize the flow fields of various interior subchannels in a 61-pin wire-wrapped hexagonal fuel bundle prototypical for a sodium fast reactor. The objective was to generate high spatiotemporal velocity field data for computational fluid dynamics turbulence model validation. The experimental facility employed the matched-index-of-refraction and modern laser-based optical measurement techniques. It is the largest transparent hexagonal test fuel assembly. Measurements were performed in two planes parallel to the axial flow, Interior-1 and Center-2. The Interior-1 location captured fluid interactions in four narrower subchannels formed by the exterior row of pins near the hexagonal duct wall. The Center-2 location captured fluid interactions in two wider subchannels spanning from the center pin out to the hexagonal duct wall. All measurements have been performed at a Reynolds number of 19,000. Results include discussion about statistical convergence of the dataset, along with flow statistics such as ensemble-averaged velocity, root-mean-square fluctuating velocity, Reynolds stress, and integral length scales. [ABSTRACT FROM AUTHOR]

The development of NekRS, a GPU-oriented thermal-fluids simulation code based on the spectral element method (SEM) is described. For performance portability, the code is based on the open concurrent compute abstraction and leverages scalable developments in the SEM code Nek5000 and in libParanumal, which is a library of high-performance kernels for high-order discretizations and PDE-based miniapps. Critical performance sections of the Navier–Stokes time advancement are addressed. Performance results on several platforms are presented, including scaling to 27,648 V100s on OLCF Summit, for calculations of up to 60B grid points (240B degrees-of-freedom). • We describe the development of NekRS, a GPU-oriented spectral element code for CFD. • NekRS merges high-performance libParanumal kernels with the scalability of Nek5000. • Portability is realized through the OCCA kernel abstraction coupled with MPI. • Performance studies are presented on several platforms, including all of Summit. [ABSTRACT FROM AUTHOR]

• Multiphysics multiscale gas cooled pebble bed reactor model using NEAMS tools. • Generation of local and global point reactivity coefficients using a coupled neutronics thermal fluids model. • Comparison of local and global reactivity coefficients during a load-following transient scenario. The systems analysis of anticipated operating occurrences and design basis accidents for pebble-bed reactor systems requires knowledge of neutron point kinetics equations (PKE) parameters. Typically, the generation of PKE parameters is performed in a global manner using standalone neutronics calculations, without the inclusion of thermal fluid distributions. We utilize Griffin and Pronghorn for generating global and local PKE parameters which includes the use of thermal fluid distributions to account for localized effects. This work establishes a methodology for calculating PKE parameters for a pebble bed reactor with a coupled neutronics/thermal fluids analysis. PKE parameters generated on a global and local basis are compared against a diffusion solve for a typical load-following transient. Locally-generated neutron kinetic parameters are able to reduce the maximum error in the transient power level from 5% to below 1.5%; along with this, bulk temperature errors were reduced from 12 K to 4 K. [ABSTRACT FROM AUTHOR]

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a prototype of a Generation IV reactor that will be constructed in Mol, Belgium. It is a liquid metal fast reactor, using lead-bismuth eutectic (LBE) as a coolant. Flow-induced vibrations resulting in fretting wear could be of concern due to the high density of LBE. Additionally, a wire is wrapped around each fuel rod to preserve their mutual distance in the hexagonal array, which affects the flow pattern. In this research turbulence-induced vibrations are investigated. Large-eddy simulation is a sound technique to acquire turbulent loads, but it is computationally demanding to combine this in a two-way coupled fluid–structure interaction (FSI) simulation. However, because accurate prediction of the vibration of cylinders subject to liquid flow typically requires a two-way simulation due to the added-mass effect, a methodology consisting of multiple steps is proposed. First, the modal characteristics of the pin are determined in a fully coupled FSI simulation of a bare bundle. The material properties used in this model are modified to take into account the effect of the wire, which was determined experimentally beforehand. The resulting characteristics are then used to construct a realistic structural model, employed in a one-way simulation with turbulent loads from a large-eddy simulation of a wire-wrapped bundle. In this structural model contact with neighboring rods and the hexagonal duct is detected, providing realistic boundary conditions to the rod and allowing to extract contact forces. These forces are further post-processed together with the sliding displacement to calculate work rates, which is an important parameter for fretting wear prediction. Two pins of a 19-pin bundle were investigated, the central pin and a pin located in one of the corners. The simulation is repeated for multiple friction coefficients, and its impeding effect of friction on sliding motions was confirmed to moderate fretting. • Structural model tuned to modal properties from two-way FSI simulation of bundle. • LES results of wire-wrapped bundle processed into load history exciting vibrations. • Contact detected in structural model, yielding contact forces and displacements. • Results post-processed into work rates and estimation of fretting wear. [ABSTRACT FROM AUTHOR]

• First-of-a-kind HCF estimation for SFR with high fidelity codes was demonstrated. • The results indicate the reduction of the HCF due to less modeling uncertainties. • The tool can be applied to different reactor design. The evaluation of hot channel factor (HCF) is of great significance to the quantification of safety margins for reactor designs. In this paper, HCFs for a sodium-cooled fast reactor (SFR) are evaluated with the Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) toolkit, which is developed under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign of DOE for multi-physics reactor performance and safety simulations. The high-fidelity neutronics and thermal hydraulics solvers PROTEUS and Nek5000 in the SHARP toolkit are coupled to perform the multi-physics simulations for HCF evaluation. The HCFs induced by cladding manufacturing tolerance, fissile content mal-distribution, wire orientation and uncertainties on the cladding, coolant, and fuel properties are evaluated for a reference core SFR design (AFR-100). The HCFs calculated with the SHARP toolkit are compared to legacy HCFs for similar reactor types. The comparison demonstrates the reduction or elimination of modeling uncertainties in the calculation of HCFs using high fidelity advanced modeling and simulation tools without the need of expensive experiments. Moreover, the reduction of the uncertainties on HCFs evaluation allows an increase in nominal parameters and safety margin, which in turn improves the economic competitiveness of the SFR. [ABSTRACT FROM AUTHOR]

• CFD methods aid the design of helical coil steam generators. • Various RANS-based turbulence models are assessed against reference LES data. • Importance of resolving wall boundary in flow separation is identified. • Recently developed STRUCT with resolved wall treatment shows the best predictions. Helical-coil steam generators are being adopted in a number of advanced nuclear reactor designs because of their increased heat transfer efficiency and compactness. Due to the limited operational experience and the expected high cost of dedicated experiments, it is imperative to demonstrate the modeling capability. Computational fluid dynamics is ideally suited for this problem, due to its high resolution and accuracy. However, traditional low-cost URANS turbulence models have shown limited applicability for cross-flow in helical tube bundles. On the other hand, LES methods provide high-accuracy and high-fidelity data with prohibitively high computational cost for design use. In this work, the recently proposed hybrid second-generation model STRUCT is compared to classic URANS turbulence formulations against high-fidelity LES simulations. The STRUCT model produces accurate predictions for all relevant quantities, demonstrating high potential for reproducing helical-coil tube bundle flow phenomena accurately, at an affordable computational cost. [ABSTRACT FROM AUTHOR]

• The paper provides a review of progress with respect to validation of CFD for application to LMFR fuel assemblies. • Reference data production programs are described for 7-, 19-, 61-, 127-, and infinite pin bundles. • Additionally related numerical validation programs are explained including the achievements in these programs. • Results indicate that an accuracy within 12.5% for RANS models should be feasible for all bundle sizes and all parameters checked. • All applied thermal validation simulations used the standard Reynolds analogy giving room for improvement. Liquid metal fast reactors (LMFRs) are foreseen to play an important role in the future of nuclear energy, thanks to their increased fuel utilization and safety features profiting from the optimal heat transfer performance of the metallic coolants. Accurate thermal-hydraulic analysis of their fuel assemblies, typically employed with wire-wraps as spacers, is recognized as a crucial scientific and engineering contribution to support the deployment of such technology. This challenges the modeling and simulation community. To this aspect, various reference databases (both experimental and numerical) for different wire-wrapped fuel assembly configurations have been created recently and are being used for validation of engineering simulation approaches based on Reynolds Averaged Navier Stokes (RANS) modelling. These databases include: • 7-pin rod bundle: A detailed experiment with Particle Image Velocimetry (PIV) is performed. In order to allow accurate measurements of the flow topology, a matched-index-of-refraction technique was used employing water as working fluid. • 19-pin bundle: A series of experiments is performed covering a wide range of Reynolds and Peclet numbers as well as thermal powers. The experiments use liquid lead-bismuth eutectic as working fluid. The measurements include pressure drop and local temperatures. • 61-pin rod bundle: This large eddy simulation including conjugate heat transfer from the pin cladding to the coolant allows to bridge the gap from small bundles (less than 37 pins) to large bundles (more than 37 pins). In literature, a fundamental different behavior has been observed for small bundles compared to large bundles. • 127-pin bundle: Isothermal experiments using lead-bismuth eutectic characterizing pressure drop are performed on a full scale fuel assembly representative for the MYRRHA reactor. • Infinite pin bundle: This reference quasi-direct numerical simulation profits from periodicity in all directions. It provides a detailed view into the flow field and in addition reveals details of the heat transfer from the rod bundle into the flow. Reference databases aim to serve the nuclear scientific community to validate engineering simulation approaches. The paper will introduce these reference databases, and how they have been used to validate RANS based turbulence modelling approaches within a mainly European context. [ABSTRACT FROM AUTHOR]

An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework.