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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 1,297 records · Page 72
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

The coupled neutronics and thermo-fluid dynamics design characteristics of the XE-100 200 MW{sub TH} reactor - 18552

The Xe-100 is a high temperature helium-cooled, graphite moderated, pebble bed reactor with a 10 times through, multi-pass (MEDUL=MEhrfacDUrchLauf) fueling scheme. The design, characterized as a Generation IV reactor, is based on proven technology that has been demonstrated a number of times both experimentally and as commercial reactors. This proven foundation ensures high levels of design and technical readiness. In the overview provided of the design it is shown how the excess reactivity is limited by continuous refueling, whilst the selected core power density will ensure passive heat removal. In this way the intrinsic safety design features of a typical GEN IV design are guaranteed. It is shown that no upset event will lead to a significant amount of radiological release. The coupled core neutronic and thermo-fluid dynamics design for the Xe-100 is performed with the VSOP-A system of codes. For the equilibrium core, VSOP results demonstrate that the spherical fuel powers (maximum 1.9 kW) and operational temperatures (<900 deg. C) fall well within the envelope of the design criteria. Adequate reactivity control and long-term, cold shutdown are provided by two separate, diversely-actuated systems, while the overall negative reactivity temperature coefficient is illustrated over the total operational range. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

V and V of the HTR Code Package (HCP) as a comprehensive HTR steady state and transient safety analysis framework - 18664

The HTR code package (HCP) allows for the simulation of several safety-related aspects of a High Temperature Reactor core in a highly integrated manner. HCP currently couples the thermo-fluid dynamics and time dependent neutronics code MGT, the spectrum code TRISHA, the burn up code TNT, the fuel management code SHUFLE and the fission product release code STACY. During its development, state-of-the-art programming techniques and standards are applied. The cross sections in HCP are generated by a 0D- or 1D solver featuring an innovative approach to treat the double heterogeneity of pebble fuel. Also major improvements have been made to optimize the nuclear data library based on ENDF/B-VII. Two advanced fuel management models are provided by the code module SHUFLE that go far beyond the capabilities of existing system codes like VSOP. The source term analysis code module STACY is coupled to HCP providing release rate calculations with a high spatial resolution making use of the nuclide densities provided by TNT. An outstanding new feature of HCP is the possibility to simulate long-term operation scenarios based on OTTO or MEDUL fuel shuffling schemes as well as selected transients in one integrated code package. Even alternating steady state/transient/steady state simulations are possible. So with HCP different fuel strategies and their influence on various kinds of accidents can be examined with one consistent reactor model. This paper provides an overview of the development status of the HCP and reports about selected benchmark results. It is demonstrated that the new system code HCP is capable to replace existing stand-alone codes like VSOP, TINTE/MGT, FRESCO or PANAMA while introducing new features, which so far to our knowledge were not available in the field of HTR safety research. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Proper orthogonal decomposition analysis of parallel twin jets measured using particle image velocimetry - 18685

The mixing of turbulent jets has been found in several locations of High Temperature Gas-cooled Reactors (HTGRs) including upper and lower plenums, where jets impinge into open space from coolant channels. In the present work, a parallel twin jets experimental facility was used to generate velocity field of mixing jets by means of Particle Image Velocimetry (PIV), which can be eventually used for validation of Computational Fluid Dynamics (CFD) simulations. The facility has two fixed parallel rectangular jet nozzles (5.8 mm in width, 87.8 mm in length) submerged in the middle of the water tank, which allows an assumption of free jet mixing. In order to compare the experimental data produced to existing data, Re = 4000 {approx} 5000 at the jet outlet was selected. In the past, several statistically meaningful data have been reported using Laser Doppler Velocimetry (LDV) and Hot Wire Anemometer (HWA), which can support Reynolds Averaged Navier-Stokes turbulent models. However, recent CFD simulations using Large Eddy Simulation (LES) require high resolution experimental data in time and space. Thus, in this study, in addition to statistically enough number of data sets for RANS models validation, high temporal and spatial resolution data were collected for the validation of LES calculations. In order to study the turbulent structure Reynolds decomposition and Proper Orthogonal Decomposition (POD) were applied to the data obtained. These decomposition analyses visualized multi-scale vortices and the transportation of eddies. In terms of twin jets structure, the present experimental data showed good agreement with other techniques for the important characteristic parameters including the length of the converging region and the location of the merging point. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Tuning and Verification of an Internal Flooding Model for Dynamic PRA

In probabilistic risk assessment (PRA), the likelihoods and consequences of basic events are combined to determine possible system level outcomes and the likelihood of each outcome given an initial event. Both the consequences and likelihoods of basic events may be informed by the use of mathematical models or experimental data regarding the phenomena of interest. In either case, the impact of uncertainties on the results need to be evaluated. If a model is used to predict the evolution of an initiating event, uncertainty quantification can be computationally challenging for complex phenomena with many epistemic (e.g., model) uncertainties. In this work, a reduced order model is introduced for the flooding analysis of a multi-room building at a nuclear power plant under uncertainties which can require long run times using high fidelity computer models. The reduced order model focuses on macro-level transfer of water between rooms and tracks the water level within each room. In order to verify the model's fidelity, its results for the phenomena involved in a scenario of interest are compared to those of a computational fluid dynamics (CFD) code, which are taken to represent reality. This work does not focus on the overall fidelity of the flooding model, as the goal of the project for which it was developed is to develop a methodology rather than to analyze a particular case. The focus instead is on the tunability and verifiability of the reduced order model for relevant physical measures. In analyzing a particular case, the model may be required to incorporate more realistic description of phenomena; however, the procedure described in this work may be used to verify its performance as well. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Benchmark of Spacer Grid Models used in Nuclear Power Plants

The fuel bundles of both PWR and BWR reactors require spacers to maintain the integrity and spacing of the slender fuel rods. These spacers can provide benefits such as increased mixing and turbulence, which increases heat distribution, leading to better Departure from Nucleate Boiling (DNB) conditions in PWRs. They also enhance the critical power ratios by increasing the thickness of the liquid layer in BWRs. However, the spacers also cause a pressure drop which must be engineered into the performance and operation. The flow within a rod bundle involves a number of complicated factors, such as cross flow, (de)entrainment of liquid layer, bubble or droplet break up/join. This requires specialized 3 dimensional computational fluid dynamics for the complete flow, however, most of the newly developed models have been accomplished by industry, so they are proprietary and in addition require extensive programming and cost to implement. In the public domain, a number of 1 dimensional equations have been developed that could benefit in initial design and decision making, if the equations are valid in the new and different applications the models are used. In 2002, an OECD/NRC BFBT benchmark workshop determined the variables used in the development of models, to predict flow within nuclear fuel bundles, had not been updated with technological advances to replace general theory and characteristics. If those properties could be identified and modeled, they could be applied to a wider range of conditions. This study utilizes NUPEC BFBT tests' measured data, on an average spacer over a range of pressure and flows, which was determined to be the most detailed measured data available. The focus is to describe the mathematical models of spacer grids and validate them with experimental results. This study consists of compiling and comparing available and established single phase equations and models computations. While these models may not have the desired degree of accuracy for final decisions, they may be accurate enough for the initial phases of design. No conclusion on a universal equation that fits all boundary conditions accurately or detailed effect of properties was determined. However, by using the BFBT experimental data, boundary conditions where the spacer grid models can be useful were identified. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Neutronics and computational fluid dynamics analyses for a natural circulation loop of molten fuel salt

A conceptual reactor system of molten fuel salt loop designed for Mo-99 production was analyzed by coupling TRITON neutronics and Fluent computational fluid dynamics (CFD) simulations. The loop system operated in 1 MW thermal power features natural circulation and potential extraction of non-soluble fission products (FP) in fuel salt. The behavior of natural circulation was studied and the power and temperature distributions in equilibrium were obtained. A long-term fuel depletion analysis of the system considering fuel circulation and reprocessing was then performed using an approximate stepwise sequence of simulations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Computational Fluid Dynamics Simulation of a Pilot-Scale Waste Glass Melter

The INL has been developing a high-fidelity computational fluid dynamics (CFD) model of a Joule-heated ceramic-lined waste glass melter. The purpose is to better understand the dynamics of heat transfer within the system, in order to provide support for the optimization of the Hanford Waste Treatment Plant (WTP) that will be used for vitrification of legacy tank waste. The simulation presented here is for the DM1200 pilot-scale system, which has been operated by the Vitreous State Laboratory (VSL) over the past several decades. This system is about one-third of the scale of the full size WTP melter. Experimental data from this system is being used to validate computational simulations. The examination of the melter with a fully coupled CFD model should give an accurate assessment of the heat transfer into the reactive cold cap region. Heat is provided to the cold cap region from conduction and convection from the molten glass pool as well as radiation from the hot plenum above the cold cap. The CFD model implements a mathematical model of the cold cap developed for the assessment of the melting rate during the vitrification of tank waste. This sub-model implemented directly into the CFD model combines the mass and enthalpy balance of the cold cap together with the conversion kinetics of the feed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Flow Field and Pressure Measurements in a 61-Pin Wire-Wrapped Bundle

Texas A and M University is conducting isothermal flow tests in a wire-wrapped, 61-pin hexagonal fuel bundle to support the research on advanced nuclear fuel development sponsored by the Department of Energy (DOE). The purpose of these tests is to produce high spatial- and temporal-resolution measurements of the flow and pressure measurements at different locations of the test bundle. The data resulting from these tests will be used to benchmark numerical simulations of the experiments using both industry-standard Reynolds-averaged Navier-Stokes (RANS) based Computation Fluid Dynamics (CFD) methods and a highly resolved Large Eddy Simulation (LES) computer code. This article described the main features of the experimental facility, its capabilities in producing experimental data of pressure drop and velocity fields at different axial locations and within different exterior and interior sub-channels, using advanced flow visualization techniques. Examples of experimental data outputs and typical locations of the measurements that will included in the research scope are also shown in this article. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Pressure Measurements in a Wire-Wrapped 61-Pin Hexagonal Fuel Bundle

Texas A and M University is conducting isothermal experiments in a wire-wrapped 61-pin hexagonal fuel bundle to support the research on advanced nuclear fuel development sponsored by the Department of Energy (DOE). The purpose of these experiments is to produce high spatial and temporal resolution measurements of the flow and pressure measurements at different locations of the test bundle. The experimental data will be used to benchmark numerical simulations of the experiments using both industry-standard Reynolds-averaged Navier-Stokes (RANS) based computational fluid dynamics (CFD) methods and highly resolved large eddy simulations (LES). (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

CFD Verification and Validation of Wire-Wrapped Pin Assemblies

TerraPower is participating in a cooperative project among industry, the national lab system, and academia to perform Verification and Validation (V and V) of Computational Fluid Dynamics (CFD) methods for predicting the flow and heat transfer within liquid-metal-cooled nuclear fuel assemblies with wire-wrapped fuel pins. This project, consisting of both experimental and numerical components, uses surrogate fluids and electrically heated fuel pins to substitute for the liquid metal coolant and the nuclear heat source. The experiments include both unheated assemblies to measure the flow velocities with Laser Doppler Velocimetry (LDV) and Particle Image Velocimetry (PIV), as well as heated assemblies to measure the temperatures of the fluid and the simulated fuel pins with thermocouples. Both sets of experiments also measure the pressure at multiple locations in the system. The numerical component of this work involves high-fidelity Large-Eddy Simulation (LES) modeling and industry-standard Reynolds-Averaged Navier-Stokes (RANS) modeling of the experiments. Furthermore, the experiments and simulations will be conducted on two duct geometries: the first is a non-deformed hexagonal duct that has flat internal duct faces that represents a beginning of life geometry. The other is a deformed duct where the internal hexagonal flow area enlarges then decreases along the axial height to represent the changes that occur to the duct geometry from pressure and irradiation effects in the core. This paper describes the methods and results of the industrial, RANS-based CFD code verification, solution verification, and solution validation. The V and V simulations are being performed for the following experiment test geometries: - Isothermal non-deformed bundle; - Heated non-deformed bundle; - Isothermal deformed bundle; - Heated deformed bundle. These flows have been computed at first with the pre-test design specifications, without the aid of any experimental results. Later, post-test calculations will be performed incorporating the actual test conditions and input from some of the experimental results. These RANS simulations of helically wire-wrapped fuel assemblies employ meshes of bare pins without the wire-wrap in the bundle. The effect of wire-wrap on the flow is accounted for by introducing a momentum source into the governing fluid equations. The momentum source is only applied to cells corresponding to the location of the wire-wrap, and its components in each cell are based on the local flow field. This methodology simplifies the geometry and mesh generation and reduces the cell count, while still capturing the effects of the wire-wrap. The momentum source approach is used for all of the assembly simulations performed in this study. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Numerical Study of Vortex Core in Turbulent Swirling Flow

Swirling flow has attracted a great deal of research interest. An application of swirling flow can be found in many engineering areas; such as gas cyclone and heat transfer enhancement. In nuclear engineering, swirling flow is used as heat transfer enhancement in rod bundle by installing grid spacer and mixing vane. Because of its extensive applications, there has been a great number of experimental and numerical investigations of swirling flow. McClusky et al., experimentally studied mapping of the lateral velocity in rod bundle using Laser Doppler Velocimetry (LDV) and Conner et al. predicted the flow field in rod bundle by CFD. To understand the characteristics of swirling flow such as tangential velocity and vortex core swirling flow, some researchers investigated swirling flow in a pipe instead of in rod bundle. This is due to the similarity between swirling flow in a pipe and rod bundle. In recent study, flow mapping of 2D velocity field of single-phase swirling flow had been conducted experimentally by Particle Image Velocimetry (PIV). However, this method needs calibration and complicated. The new measurement technique for flow mapping in swirling flow was done by phased array Ultrasonic Velocity Profiler (PAUVP). Hamdani et al., successfully plotted velocity field and determined the swirling core region (vortex core). This technique is nonintrusive and no need calibration. However, there was a limitation of the measurement because phased array UVP cannot capture 360 deg. velocity field inside the pipe. In addition, tangential velocity in CFD results was not agree well with experiment data due asymmetric problem. To evaluate this limitation of experimental method, the numerical simulations using Computational Fluid Dynamics (CFD) needs to be performed. The objective of this study is to use the commercial CFD code Fluent to model the developing turbulent swirling flow induced by twisted tape inside a straight pipe, and to gain an understanding of such a flow and the turbulence models involved in the simulation. Straight pipe is used instead of rod bundle for a simplification. In this work, the Renormalized Group (RNG) k-{epsilon} turbulence model and Reynolds Stress Model (RSM) turbulence model, are performed to study the phenomena of vortex core in a pipe with twisted tape inserts. Effects of the Reynolds number and asymmetric of twisted tape (inclination, {alpha} = 5 deg. and {alpha} = 20 deg.) on swirling core position are examined using water as the testing fluid. Furthermore, the Navier-Stokes equations will be solved to obtain results that can be validated with the previous experimental results. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Large Eddy Simulations for a Helical Coil Steam Generator

The desire for scalable power plants in remote locations and specialized applications has brought forth the small modular reactor (SMR). Now, smaller light water reactors are seen as viable means for meeting global energy requirements. Generally, straight or U-shape tubes are used to exchange heat from the primary to secondary coolant; however this is not ideal when size is a criterion. Helical coil steam generators (HCSGs) are advantageous for SMRs due to their enhanced heat transfer in compact spaces. Still, there is the concern of safety as steam generator tubes could rupture due to vibrations and fatigue on the material. Understanding flow characteristics and quantifying pressures for these tubes are important components for the design of safety and efficiency in these steam generators. The current study uses the spectral element computational fluid dynamic (CFD) code, Nek5000, to develop large eddy simulations (LES) for validation against experiments that will be conducted at Texas A and M University. Time-averaged pressure data will be compared with pressure tap readings, and velocity profile with particle image velocimetry (PIV) measurements. The presented preliminary results investigate the differences between 5.-, 7.-, and 9.-order polynomial simulations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Dakota Uncertainty Quantification Methods Applied to the CFD code Nek5000

Predictive modeling and simulation of the performance of a nuclear reactor and its fuel is a challenging task because of the large number of coupled physical phenomena that must be addressed. In addition to this intrinsic complexity, model uncertainty must be accounted for in any analysis if the model will be used to facilitate design or operational decisions which may impact safety and performance. Rigorous, structured uncertainty analyses are performed by first characterizing the model's input uncertainties and, by propagating the uncertainty through the model in order to estimate the output's uncertainty. This paper presents the state of advancement of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) project to characterize the uncertainty of the CFD code Nek5000 using the Dakota package. Nek5000 is a high order spectral element computational fluid dynamics (CFD) code developed at Argonne National Laboratory for high resolution spectral-filtered large eddy simulations (LESs) and unsteady Reynolds averaged Navier-Stokes (RANS) simulations. The Dakota package developed at Sandia National Laboratory can be integrated with many scientific and engineering codes to facilitate efficient, effective uncertainty quantification (UQ) and sensitivity analyses. Dakota is designed to interface with external packages/codes such as Nek5000 through a script-driven interface called fork. Such interface provides maximum flexibility to the users for pre- and post-processing tasks preceding and following each function evaluation. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Verification and Validation of CFD Simulations involving Twin Rectangular Jets using Steady RANS in Star-CCM+

Analyses such as component design and safety analysis can be conducted through careful application of computational fluid dynamics (CFD) codes. Before these codes can be considered ready for specific analysis, benchmarking (validation) against applicable datasets is needed. The twin jet water facility (TJWF) developed originally at the University of Tennessee, Knoxville has been purposed for aiding in this type of work. The TJWF allows for large data sets to be created using state of the art non-invasive measurement techniques within a mostly transparent testing volume. Utilizing velocity data collected in the TJWF by Wang et. al., the Star-CCM+ code was benchmarked a using steady Reynolds averaged Navier Stokes (RANS) framework for a combining two jet case. Careful efforts to properly conduct verification of the numerical work was undertaken with a sample reported in this summary. The computational results were validated against the experimental work and found to be within acceptable limits. The parameters of interest were the merging and combining points (MP and CP) of the twin jets, and the velocity profiles at increasing heights after injection. Fig.1 provides a graphical depiction of the twin jets combining within a free shear volume with key geometric factors included. The MP and velocity profiles were predicted with reasonable agreement. Whereas, the combining point of the twin jets was found to be over-predicted by {approx}24% difference. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Development of CFD Flow Analysis for Nuclear Facility Ventilation Systems

Understanding and modeling of particulate transport has importance related to safety in ventilated environments that contain hot cells, gloveboxes, and fume hoods. In particular, good modeling of the airflow in these spaces is required to understand areas of stagnation where particles drop out of the free stream and accumulate. This accumulation of particulate can cause either contamination and/or higher than acceptable local dose levels. It is known that multiple sites across the DOE complex take credit for hot cells, gloveboxes, and/or hoods in their safety basis for providing a defense-in-depth benefit for both onsite and offsite releases. By providing confinement of radioactive materials, such features serve to reduce direct doses to facility workers and mitigate the consequences to the environment due to an uncontrolled release. Each of these features has access points that interface with the personnel space. Understanding how air flow behaves at these access points is of great interest to those performing hazard analysis. Computational Fluid Dynamics (CFD) can be used to analyze and model the flows in hot cells and glove boxes as a way to confirm operation. While CFD capability continues to advance, there are still important modeling assumptions that are left to the analyst's discretion. These are most notably the choice of the turbulence models, mesh structure, and wall boundary condition assumptions used in the model. In this work, we apply CFD in modeling airflow through an occupiable work area of the Alpha Gamma Hot Cell Facility (AGHCF) at Argonne National Laboratory (ANL). This modeling is then compared to field test data to both inform and verify the modeling assumptions. Based on the results of field tests, modeling assumptions and boundary conditions are refined and the process is repeated until the results are found to be reliable with a high level of confidence.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Evaluation of the shear rate effect on the interfacial forces acting on a single bubble

Owing to the outstanding characteristics of mass and heat transfer efficiency, two-phase bubbly flows have wide applications in the nuclear engineering fields. However, the physics behind the bubbly flows phenomenon can be very complex due to the various parameters and phenomena involved. Thanks to the increasing computing power and advanced algorithms, computational fluid dynamics (CFD) approach has emerged as an effective tool for exploring the bubbly flows. In typical laminar channel or pipe flow, the flow field develops a parabolic velocity profile due to the wall friction. Assuming the shear rate across the bubble diameter is constant, the bubble will encounter various shear rates depending on its transverse position. Different shear rates may lead to the acceleration of transverse migration speed and influence the bubble rising velocity. In this summary, we evaluate the effect of shear rate on the interfacial forces (lift and drag) acting on a single bubble with the shear rate ranging from 3.8 to 8.3 s{sup -1} and the Eotvos number ranging from 2.58 to 4.45.

42 ENGINEERING↗
DOE OSTI2017

CFD Simulation of Bubble Growing Under a Downward Facing Surface

The heat removing of downward facing surface is an important issue in many areas, including nuclear safety, solar storage system and electronic cooling devices. Therefore, the characteristic and heat removal capability of downward facing boiling were investigated by many experiments in recent years. Among these studies, the thickness of bubble is considered as an important parameter. The increasing of bubble thickness will increase the thermal resistance of cooling process and reduce the cooling ability. In order to focus on the dynamics of bubble thickness, Das et al. injected air into sodium dodecyl sulphate (SDS) solution through a nozzle under the horizontal downward facing surface. The bubble will be formed by the air injection and kept under the horizontal plate. The diameter of nozzle, horizontal plate and SDS solution tank are 2 mm, 20 mm and 30 mm, respectively. The growing processing and the stable thickness of the bubble were recorded in the experiment. Basing on the conditions of Das's experiment, the appropriate computational fluid dynamics (CFD) simulation was conducted. The growing process and the stable thickness of bubble under difference contact angle are predicted by CFD in this study.

42 ENGINEERING↗
DOE OSTI2017

Time-Resolved PIV/PTV Measurements on Interior Subchannels of a Wire-Wrapped 61-pin Hexagonal Fuel Bundle

Liquid metal fast reactor fuel bundles using sodium as a coolant typically utilize a tightly packed triangular lattice of fuel pins enclosed in a hexagonal duct. Often, the fuel pins are helically wrapped with a wire spacer. The primary function of the wire spacers is to maintain a gap between adjacent fuel pins. They also enhance sub-channel mixing to increase convective heat transfer and mitigate vortex-induced vibration. Texas A and M University has conducted isothermal flow experiments in a wire-wrapped 61-pin hexagonal fuel bundle to support the research on advanced nuclear fuel development sponsored by the Department of Energy (DOE). These tests produced high spatial and temporal resolution measurements of the velocity field and pressure at different locations in the fuel bundle. The experimental facility employs matched-index-of-refraction (MIR) techniques and laser diagnostic velocity measurement techniques such as particle image velocimetry (PIV), particle tracking velocimetry (PTV), and laser Doppler velocimetry (LDV). The result of this experimental activity will be an experimental database of pressure and velocity field measurements that will be suitable for validating computational fluid dynamics (CFD) methods, such as Reynolds-averaged Navier-Stokes (RANS) and highly resolved large eddy simulation (LES). PIV results on the set of edge subchannels near the hexagonal duct have been presented in another paper. This paper presents PIV results on two sets of interior subchannels, defined as Interior- 1 and Center-2, at a Reynolds number (Re) of 19,000 in the turbulent regime.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗