Relevant bibliographies by topics / Wire-wrapped rod bundles

Academic literature on the topic 'Wire-wrapped rod bundles'

Author: Grafiati
Published: 10 December 2022
Last updated: 28 January 2023

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Journal articles on the topic "Wire-wrapped rod bundles"

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Ninokata, Hisashi, Apostolos Efthimiadis, and Neil E. Todreas. "Distributed resistance modeling of wire-wrapped rod bundles." Nuclear Engineering and Design 104, no. 1 (October 1987): 93–102. http://dx.doi.org/10.1016/0029-5493(87)90306-2.

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Suh, K. Y., N. E. Todreas, and W. M. Rohsenow. "Mixed Convective Low Flow Pressure Drop in Vertical Rod Assemblies: II—Experimental Validation." Journal of Heat Transfer 111, no. 4 (November 1, 1989): 966–73. http://dx.doi.org/10.1115/1.3250812.

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An experimental study has been conducted to validate the predictive models and correlations for laminar and transition flow frictional pressure loss in vertical rod bundles under mixed convection conditions. An experimental procedure has been developed to measure low differential pressures under mixed convection conditions in 19 heated rod bare and wire-wrapped assemblies. The proposed model has been found successfully to predict the effects of wire wrapping, power skew, transition from laminar regime, developing and Interacting flow redistributions, and rod number on the friction loss characteristics in bundle geometries over the bundle average Grq/Re number range of 6 to 18,500.
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Dix, Adam, and Seungjin Kim. "A novel friction factor model for wire-wrapped rod bundles." Nuclear Engineering and Design 401 (January 2023): 112104. http://dx.doi.org/10.1016/j.nucengdes.2022.112104.

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Bovati, Octavio, Mustafa Alper Yildiz, Yassin Hassan, and Rodolfo Vaghetto. "Pressure drop and flow characteristics in partially blocked wire wrapped rod bundles." Annals of Nuclear Energy 165 (January 2022): 108671. http://dx.doi.org/10.1016/j.anucene.2021.108671.

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Choi, Sun Rock, Hyungmo Kim, Seok-Kyu Chang, Hae Seob Choi, Dong-Jin Euh, Hyeong-Yeon Lee, and Won Sik Yang. "Assessment of subchannel flow mixing coefficients for wire-wrapped hexagonal fuel rod bundles." Annals of Nuclear Energy 166 (February 2022): 108810. http://dx.doi.org/10.1016/j.anucene.2021.108810.

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Bovati, Octavio, Mustafa Alper Yildiz, Yassin Hassan, and Rodolfo Vaghetto. "RANS simulations for transition and turbulent flow regimes in wire-wrapped rod bundles." International Journal of Heat and Fluid Flow 90 (August 2021): 108838. http://dx.doi.org/10.1016/j.ijheatfluidflow.2021.108838.

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Carajilescov, Pedro, and Elói Fernandez y Fernandez. "Model for subchannel friction factors and flow redistribution in wire-wrapped rod bundles." Journal of the Brazilian Society of Mechanical Sciences 21, no. 4 (December 1999): 589–99. http://dx.doi.org/10.1590/s0100-73861999000400003.

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Hu, Rui, and Thomas H. Fanning. "A momentum source model for wire-wrapped rod bundles—Concept, validation, and application." Nuclear Engineering and Design 262 (September 2013): 371–89. http://dx.doi.org/10.1016/j.nucengdes.2013.04.026.

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Kim, Hansol, Yu Min Chen, and Yassin Hassan. "Prediction of pressure drop in hexagonal wire-wrapped rod bundles using artificial neural network." Nuclear Engineering and Design 381 (September 2021): 111365. http://dx.doi.org/10.1016/j.nucengdes.2021.111365.

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Chen, S. K., Y. M. Chen, and N. E. Todreas. "The upgraded Cheng and Todreas correlation for pressure drop in hexagonal wire-wrapped rod bundles." Nuclear Engineering and Design 335 (August 2018): 356–73. http://dx.doi.org/10.1016/j.nucengdes.2018.05.010.

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Dissertations / Theses on the topic "Wire-wrapped rod bundles"

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Kindfuller, Vincent John. "Improvements to the Cheng-Todreas wire-wrapped rod bundle friction factor correlation in response to pin number and in the transition flow region." Thesis, Massachusetts Institute of Technology, 2016. http://hdl.handle.net/1721.1/106697.

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Thesis: S.B., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2016.
This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.
Cataloged from student-submitted PDF version of thesis.
Includes bibliographical references (pages 51-52).
The Cheng-Todreas Detailed (CTD) Pressure Drop Correlation (1986) is the most accurate correlation for measuring the pressure drop of sodium coolant through a Sodium-Cooled Fast Reactor (SFR). Although CTD is the most accurate correlation, there is room for improvement with new data and modern ways to visualize data and check the accuracy of changes easily. This thesis attempts to offer a method for altering the CTD correlation to better account for changing pin numbers in SFR assemblies, and a better fit for the correlation to the data in the transition flow region between turbulent and laminar flow. Although CTD is more accurate than other correlations, it shows an inverse response to changing pin number in some geometries of bundle assemblies. In this thesis, a method is laid out to attempt to correct for that inverse response. Although no successful conclusion was reached, the thesis also offers a method for future attempts at improvement. In addition, a set of Matlab codes are offered that allow changes to be easily attempted and checked for validity. In addition, examining the data points of bundles in the transition flow regime shows possibilities for improving the accuracy of the correlation in that flow region. Two changes are implemented in this thesis: a change to the equation for the boundary between laminar and transition flow, and a change to the transition region friction factor equation. Both changes, when implemented, offer slight improvements to the overall accuracy and precision of the Cheng-Todreas Pressure Drop Correlation..
by Vincent John Kindfuller.
S.B.
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Conference papers on the topic "Wire-wrapped rod bundles"

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Brockmeyer, Landon, Lane Carasik, Elia Merzari, and Yassin Hassan. "CFD Investigation of Wire-Wrapped Fuel Rod Bundle Inner Subchannel Behavior and Dependency on Bundle Size." In 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-60831.

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Sodium fast reactor designs often implement a hexagonal array of fuel rods with wire-wrappers to encourage the exchange of coolant between subchannels. The ability to accurately predict inter-subchannel mixing can be used as a metric for turbulence model performance in capturing wire-wrapped fuel rod bundle flow behavior. In this study inter-subchannel mixing predictions by Large Eddy Simulation (LES) and Reynolds Averaged Navier Stokes (RANS) models are compared. The results indicate that the lower order RANS approach is capable of predicting inter-subchannel mixing inside a 19 rod bundle with acceptable accuracy. The RANS model was extended to 37, 61, and 91 rod bundles to observe the effects of bundle size on inter-subchannel exchange for the center-most subchannels. Transverse velocity magnitude and mass exchange were observed to increase with larger bundle sizes. Inter-subchannel mixing is observed to be a strong function of bundle size for bundles up to 91 rods. The results indicate that the inner subchannels of larger bundles may converge upon a characteristic flow pattern. The 91 rod bundle is not large enough to isolate the inner subchannels from shroud effect, and larger bundles will need to be investigated.
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Heidet, F., and S. Yoon. "Evaluation of Pressure Drop Correlations for the Wire-wrapped Rod Bundles." In Transactions - 2020 Virtual Conference. AMNS, 2020. http://dx.doi.org/10.13182/t122-31608.

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Heidet, F., and S. Yoon. "Evaluation of Pressure Drop Correlations for the Wire-wrapped Rod Bundles." In Transactions - 2020 Virtual Conference. AMNS, 2020. http://dx.doi.org/10.13182/t31608.

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Wu, Jie, and Jiejin Cai. "Initial Wire-Wrapped Angle Optimization Analysis of a Liquid Lead-Bismuth Cooled Fuel Assembly Based on OpenFOAM." In 2022 29th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2022. http://dx.doi.org/10.1115/icone29-92097.

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Abstract The helical wire spacer of liquid lead-bismuth cooled fast reactor (LFR) fuel assembly plays a significant role in the strengthening of the flow and heat transfer. However, most LFRs have a fixed initial angle of wire-wrapped direction in rod bundles, and the optimization analysis of their angle in the subchannel is absent. More analysis of the impacts that the wire spacer has on the liquid lead-bismuth eutectic (LBE) coolant should be obtained. In this paper, three different turbulence viscosity models and two constant turbulent Prandtl numbers were applied in the open source Computational Fluid Dynamics (CFD) platform OpenFOAM. The numerical results, which were considered with an extensive mesh sensitivity study, were validated against a series of experiment data. The simulations about key thermal hydraulic parameters such as temperature, velocity distribution, pressure drop, local and average Nusselt number were carried out based on 7-pin wire-wrapped rod bundles whose wires winding from internal, edge and corner channel. The results show that the k-ε model with Prt = 2.0 can be used to predict the flow and heat transfer characteristics of LBE. The influence of wire-wrapped starting position indicates that an optimum point exists in the internal channel. This work is useful in future safety design of fuel assemblies in the LFR.
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Smith, Jeffrey G., Bruce R. Babin, W. David Pointer, and Paul F. Fischer. "Effects of Mesh Density and Flow Conditioning in Simulating 7-Pin Wire Wrapped Fuel Pins." In 16th International Conference on Nuclear Engineering. ASMEDC, 2008. http://dx.doi.org/10.1115/icone16-48306.

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In response to the goals outlined by the U.S. Department of Energy’s Global Nuclear Energy Partnership program, Argonne National Laboratory has initiated an effort to create an integrated multi-physics multi-resolution thermal hydraulic simulation tool package for the evaluation of nuclear power plant design and safety. As part of this effort, the applicability of a variety of thermal hydraulic analysis methods for the prediction of heat transfer and fluid dynamics in the wire-wrapped fuel-rod bundles found in a fast reactor core is being evaluated. The work described herein provides an initial assessment of the capabilities of the general purpose commercial computational fluid dynamics code Star-CD for the prediction of fluid dynamic characteristics in a wire wrapped fast reactor fuel assembly. A 7-pin wire wrapped fuel rod assembly based on the dimensions of fuel elements in the concept Advanced Burner Test Reactor [1] was simulated for different mesh densities and domain configurations. A model considering a single axial span of the wire wrapped fuel assembly was initially used to assess mesh resolution effects. The influence of the inflow/outflow boundary conditions on the predicted flow fields in the single-span model were then investigated through comparisons with the central span region of models which included 3 and 5 spans. The change in grid refinement had minimal impact on the inter-channel exchange within the assembly resulting in roughly a 5 percent maximum difference. The central span of the 3-span and 5-span cases exhibits much higher velocities than the single span case,, with the largest deviation (15 to 20 percent) occurring furthest away from the wire spacer grids in the higher velocity regions. However, the differences between predicted flow fields in the 3-span and 5-span models are minimal.
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Pucciarelli, Andrea. "Results of a LES Application to LBE Turbulent Flow in a Wire-Wrapped Single Rod Channel." In 2021 28th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2021. http://dx.doi.org/10.1115/icone28-64153.

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Abstract Innovative Liquid Metal Fast Reactors are among the most promising concepts for the upcoming Nuclear Gen IV; nevertheless, several challenges still have to be overcome before achieving a sufficiently mature understanding of all the involved phenomena. Concerning Thermal-Hydraulics aspects, the correct estimation of the turbulent heat flux contributions and their impact on the predicted heat transfer is still being investigated. Particularly, this becomes relevant when addressing complex geometries, such as the wire-wrapped rod bundles considered for the reactor core, in which simple assumptions may no more be sufficient for a suitable representation of the actual turbulent quantities’ distributions. The present paper reports on the results of a LES application to a wire-wrapped single rod channel performed adopting the STAR-CCM+ code. The considered fluid is Lead-Bismuth Eutectic Alloy (LBE), often adopted as the working fluid in several experimental facilities built to support the development of Gen. IV LFRs while the considered Reynolds number is in the range of 8600 and the heated length is 0.1 m. The obtained results provide relevant information to be considered for the development and tuning of RANS turbulence models specialized in addressing operating conditions involving liquid-metals: the distributions of the turbulent heat fluxes (THF) is here particularly taken into account.
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Guo, Jinsong, Xueyuan Zhang, Haiqi Zhao, Daogang Lu, and Yuhao Zhang. "Three-Dimensional Numerical Simulation of the Natural Circulation Characteristics Based on PLANDTL-DHX for Different Modeling Methods of the Core." In 2022 29th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2022. http://dx.doi.org/10.1115/icone29-92364.

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Abstract The passive decay heat removal system based on natural circulation can passively remove the heat from the core, which greatly improves the safety of the nuclear reactor. The Plant Dynamics Test Loop (PLANDTL-DHX) experimental facility can simulate the flow and heat transfer characteristics of the pool-type sodium-cooled fast reactor with an independent decay heat removal system in a natural circulation state. However, the natural circulation experiments based on the PLANDTL-DHX facility are difficult to present the detailed flow characteristics in the core completely. So it is necessary to adopt numerical simulation analysis to obtain the flow characteristics in the core. While due to the complex structure of the core with wrapped wire bundles, the modeling and calculation of the pool-type fast reactor need very rich computing resources. To reduce the demand for computing resources, the model can be simplified to some extent. In this study, two modeling methods are adopted for the core: 1. The model of the rod bundles and wrapped wires are simplified by the porous media model; 2. The wrapped wires are simplified by the porous media model, while the rod bundles are retained. The PLANDTL-DHX experimental facility modeled by two different core modeling methods is numerically simulated. By analyzing and comparing the experimental data of PLANDTL-DHX, the feasibility of two different modeling methods for numerical simulation research is verified. By analyzing and comparing the calculation results of two different modeling methods, the flow characteristics in the core during natural circulation are also obtained, and the characteristics of different modeling methods are summarized. This work can provide a reference for the safety analysis and simulation calculation of pool-type sodium-cooled fast reactor.
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Jeong, Jae-Ho. "Assessment of RANS Based CFD Methodology Using JAEA Fuel Assembly Experiment." In 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-61038.

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This paper presents an assessment results for the developed RANS (Reynolds Averaged Navier-Stokes simulation) based CFD (Computational Fluid Dynamics) methodology applicable to real scale 217-pin wire wrapped fuel assembly of the KAERI (Korea Atomic Energy Research Institute) PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor). Complicated and vortical flow phenomena in the wire-wrapped fuel bundles were captured by a shear stress transport (SST) turbulence model, and by a vortex structure identification technique based on the critical point theory. The CFD results show good agreement with the JAEA experiment with the 127-pin wire-wrapped fuel assembly. The JAEA experiment study was implemented using water for validating pressure drop formulas in ASFRE code. The edge vortex structures are longitudinally developed, and have a higher axial velocity than corner vortex structures and wakes nearby pins and wires. The wire spacers locally induce a tangential flow by up to about 16 % of the axial velocity. The tangential flow in the corner and edge sub-channels is much stronger than that in the interior subchannels. The large-scale edge vortex structures have higher turbulence intensity and lower vorticity than the small-scale wakes. The corner vortex structures have lower turbulence intensity and vorticity than the small-scale wakes. The driving forces in the X-, Y-, and Z-directions are not only dependent on the axial velocity, but also significantly dependent on the angular position between the wire-spacer and rod, and the relative position between the wire-spacer and duct wall.
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Li, Minggang, Jun Wang, Changhua Nie, Xiao Yan, Yanping Huang, Zumao Yang, and Feng Xie. "Numerical Investigation of Flow and Heat Transfer Characteristics in Wire-Wrap Tight Lattice 19-Rod Bundle." In 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-15832.

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Flow and heat transfer characteristics in wire-wrap tight lattice rod bundle have been investigated through CFD code ANSYS CFX 13.0. The bundle consists of 19 fuel rods with triangular tight lattice configuration. The rod ratio of rod pitch to rod diameter is 1.167. Four wires with a diameter of 0.5 mm are helically wrapped on the surface of each fuel rod. The ratio of wire-wrap helical pitch to the rod diameter is varied from 27.5 to 52.5. Through simulating wire-wrap 3-rod bundle with tetrahedron and hexahedron grid systems, the grid system which applies to simulating the wire-wrap tight lattice rod bundle has been obtained. The predicted results of eddy viscosity based turbulence models (k–ε, SST) and Reynolds stress turbulence models (BSL, SSG) are compared with each other and several experimental correlations for friction factor and Nusselt number. The predicted results of all the turbulence models are almost the same in some respects, but the friction factor predicted by the eddy viscosity models is higher than that predicted by the RSM. The effect of wire-wrap on pressure drop, friction factor, secondary flow, heat transfer, velocity distribution and temperature distribution in different subchannels (interior, edge and corner) has been analyzed by comparing with those of the bare rod bundle. The effect of wire-wrap pitch on the flow and heat transfer characteristics has also been studied.
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McCreery, Glenn E., Hugh M. McIlroy, Kurt D. Hamman, and Hongbin Zhang. "Design of Wire-Wrapped Rod Bundle Matched Index-of-Refraction Experiments." In 16th International Conference on Nuclear Engineering. ASMEDC, 2008. http://dx.doi.org/10.1115/icone16-48585.

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Experiments will be conducted in the Idaho National Laboratory (INL) Matched Index-of-Refraction (MIR) Flow Facility [1] to characterize the three-dimensional velocity and turbulence fields in a wire-wrapped rod bundle typically employed in liquid-metal cooled fast reactors and to provide benchmark data for computer code validation. Sodium cooled fast reactors are under consideration for use in the U.S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP) program. The experiment model will be constructed of quartz components and the working fluid will be mineral oil. Accurate temperature control (to within ± 0.05 °C) matches the index-of-refraction of mineral oil with that of quartz and renders the model transparent to the wavelength of laser light employed for optical measurements. The model will be a scaled 7-pin rod bundle enclosed in a hexagonal canister. Flow field measurements will be obtained with a LaVision 3-D particle image velocimeter (PIV) and complimented by near-wall velocity measurements obtained from a 2-D laser Doppler velocimeter (LDV). These measurements will be used as benchmark data for computational fluid dynamics (CFD) validation. The rod bundle model dimensions will be scaled up from the typical dimensions of a fast reactor fuel assembly to provide the maximum Reynolds number achievable in the MIR flow loop. A range of flows from laminar to fully-turbulent will be available with a maximum Reynolds number, based on bundle hydraulic diameter, of approximately 22,000. The fuel pins will be simulated by 85 mm diameter quartz tubes (closed on the inlet ends) and the wire-wrap will be simulated by 25 mm diameter quartz rods. The canister walls will be constructed from quartz plates. The model will be approximately 2.13 m in length. Bundle pressure losses will also be measured and the data recorded for code comparisons. The experiment design and preliminary CFD calculations, which will be used to provide qualitative hydrodynamic information, are presented in this paper.
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